WorldWideScience

Sample records for irradiation embrittlement effects

  1. Irradiation embrittlement mitigation

    Torronen, K.; Pelli, R.; Planman, T.; Valo, M.

    1993-01-01

    Mitigation methods for reducing the irradiation damage on pressure vessel materials are reviewed: load leakage loading schemes are commonly used in PWRs to mitigate reactor pressure vessel embrittlement; dummy assemblies have been applied in WWER 440-type and in some old western power plants, when exceptional fast embrittlement has been encountered; shielding of the pressure vessel has been developed, but is not in common use; pre-stressing the pressure vessel has been proposed for preventing PTS failures, but its applicability is not yet demonstrated. The large number of successful annealing treatments performed in WWER 440 type reactors as well as research on the effects of annealing treatments suggest applications for western PWRs. The emergency core cooling systems have been modified in WWER 440-type reactors in connection with other mitigation measures. (authors). 37 refs., 18 figs., 2 tabs

  2. Irradiation embrittlement mitigation

    Torronen, K; Pelli, R; Planman, T; Valo, M [Technical Research Centre of Finland, Jyvaeskylae (Finland). Combustion and Thermal Engineering Lab.

    1994-12-31

    Mitigation methods for reducing the irradiation damage on pressure vessel materials are reviewed: load leakage loading schemes are commonly used in PWRs to mitigate reactor pressure vessel embrittlement; dummy assemblies have been applied in WWER 440-type and in some old western power plants, when exceptional fast embrittlement has been encountered; shielding of the pressure vessel has been developed, but is not in common use; pre-stressing the pressure vessel has been proposed for preventing PTS failures, but its applicability is not yet demonstrated. The large number of successful annealing treatments performed in WWER 440 type reactors as well as research on the effects of annealing treatments suggest applications for western PWRs. The emergency core cooling systems have been modified in WWER 440-type reactors in connection with other mitigation measures. (authors). 37 refs., 18 figs., 2 tabs.

  3. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  4. Current understanding of the effects of enviromental and irradiation variables on RPV embrittlement

    Odette, G.R.; Lucas, G.E.; Wirth, B.; Liu, C.L.

    1997-01-01

    Radiation enhanced diffusion at RPV operating temperatures around 290 degrees C leads to the formation of various ultrafine scale hardening phases, including copper-rich and copper-catalyzed manganese-nickel rich precipitates. In addition, defect cluster or cluster-solute complexes, manifesting a range of thermal stability, develop under irradiation. These features contribute directly to hardening which in turn is related to embrittlement, manifested as shifts in Charpy V-notch transition temperature. Models based on the thermodynamics, kinetics and micromechanics of the embrittlement processes have been developed; these are broadly consistent with experiment and rationalize the highly synergistic effects of most important irradiation (temperature, flux, fluence) and metallurgical (copper, nickel, manganese, phosphorous and heat treatment) variables on both irradiation hardening and recovery during post-irradiation annealing. A number of open questions remain which can be addressed with a hierarchy of new theoretical and experimental tools

  5. Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

    2005-06-01

    This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

  6. The flow effect in the irradiation embrittlement in pressure vessel steels of nuclear power plants

    Kempf, Rodolfo A.; Cativa Tolosa, Sebastian; Fortis, Ana M.

    2009-01-01

    This paper deals with the advances in the study of the mechanical behavior of the Reactor Pressure Vessel steels under accelerate irradiations. The objective is to study the effect of lead factors on the interpretation of the mechanisms that induced the embrittlement of the RPV, like those of the reactors Atucha II and CAREM. It is described a device designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. It is presented also an automatic digital image processing technique for partitioning Charpy fracture surface into regions with a clear physical meaning and appropriate for the work in hot cells. The aim is to obtain the fracture behavior of irradiated specimens with different lead factors in the range of high fluencies and to know the dependence with the composition of the alloy and with the diffusion of other alloy elements. (author)

  7. Survey of irradiation embrittlement effects on the mechanical properties of alloyed steels

    Gillemot, F.

    1992-01-01

    In the everyday engineering practice the neutron irradiation embrittlement of the PWR wall materials is measured by empirical methods like Charpy impact testing. New developments in fracture mechanics are given better material characteristics. The use of Absorbed Specific Fracture Energy Measured on tensile bars is a promising way to solve the problem. On the other hand the IAEA runs coordinated research program to correlate the chemical analysis with the rate of the neutron embrittlement. Better understanding of the physics of neutron embrittlement should help the life time management of the PWR vessels

  8. FP7 project LONGLIFE: Treatment of long-term irradiation embrittlement effects in RPV safety assessment

    May, J.; Hein, H.; Altstadt, E.; Bergner, F.; Viehrig, H.W.; Ulbricht, A.; Chaouadi, R.; Radiguet, B.; Cammelli, S.; Huang, H.; Wilford, K.

    2012-01-01

    The increasing age of European Nuclear Power Plants (NPPs) and envisaged extensions of plant lifetimes from 40 up to 80 years require an improved understanding of ageing phenomena of RPV components. The Network of Excellence NULIFE (Nuclear Plant Life Prediction) has been established to advance the safe and economic long-term operation (LTO) of NPPs by facilitating increased co-operation for applied R and D amongst members of the European nuclear community. The accurate prediction and management of RPV neutron irradiation embrittlement connected with long-term operation is an important aspect of this co-operation. Phenomena that might become important at high neutron fluences (such as flux effects and late blooming effects) have to be considered adequately in safety assessments. However, the surveillance database for prolonged irradiation times and low neutron fluxes is sparse. Consequently, there are significant uncertainties in the treatment of long-term irradiation effects. Therefore, the project LONGLIFE (Treatment of long-term irradiation embrittlement effects in RPV safety assessment) was initiated under the Euratom 7th Framework Programme of the European Commission as an umbrella project of NULIFE. LONGLIFE aims at 1) improved understanding of long-term irradiation phenomena that might compromise RPV integrity, and thereby the LTO of European NPPs, and 2) assessment of the adequacy of current prediction tools, codes, standards and surveillance guidelines for supporting long-term RPV operation. The scope of the work comprises the analysis of LTO boundary conditions; microstructural investigations and supplementary mechanical tests on RPV steels, including RPV steels from decommissioned plants; training activities; and elaboration of recommendations for RPV materials assessment and embrittlement surveillance under LTO conditions. A key part of the technical work is the selection of relevant materials for examination, e.g. which contain different weld and base

  9. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  10. Irradiation embrittlement of pressure vessel steels

    Brumovsky, M.; Vacek, M.

    1975-01-01

    A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

  11. The effect of deformation twinning on irradiation embrittlement in iron single crystals

    Kayano, Hideo; Tokutomi, Shoichiro; Yajima, Seishi; Takaku, Hiroshi.

    1978-01-01

    Single crystals of iron with the [100] crystal orientation were irradiated in JMTR with fast neutrons to a fluence of 8 x 10 18 n/cm 2 (E > 1 MeV). All samples were deformed in tension at temperatures from liquid nitrogen temperature to 200 0 C at different strain rates using an Instron-type tensile testing machine. Scanning electron microscopy of the fractured surfaces revealed that deformation twinning is difficult to occur in irradiated samples, and also that twins formed in both irradiated and unirradiated samples inhibit fracture nucleation and growth. From the results of tensile deformation of the irradiated samples deformed in tension a different strain rates at 159 0 K, it is conceived that twinning suppression is greater in the irradiated than in the unirradiated samples, and that the nucleation and growth of twins are not necessarily related to those of cracks. It is suggested that the irradiation-induced defects impede plastic deformation of the crystals and deformation twinning is suppressed by irradiation, thus causing the irradiation embrittlement. (auth.)

  12. Fluence-rate effects on irradiation embrittlement and composition and temperature effects on annealing/reirradiation sensitivity

    Hawthorne, J.R.; Hiser, A.L.

    1988-01-01

    Recent MEA investigation on the effect of neutron fluence rate on radiation-induced embrittlement accrual and the contributions of metallurgical variables to postirradiation annealing and re-irradiation behavior are reviewed. Studies of fluence-rate effects involved experiments in the UBR test reactor and separately, radiation sensitivity determinations for the decommissioned Gundremmingen (KRB-A) vessel material. Annealing-reirradiation studies employed 399 0 C and 454 0 C heat treatments. Material composition is shown to play a major role in postirradiation annealing recovery. Results illustrate effects of variable copper and variable nickel contents on recoveray of steel plate having low phosphorus levels. Composition effects on recovery were also observed for prototypic welds depicting high/low copper and high/low nickel contents and three flux types. The welds, in addition, indicate major differences in re-irradiation sensitivity. The UBR investigations revealed a significant difference in fluence rate sensitivity between the ASTM A 302-B reference plate and a submerged-arc (S/A) Linde 80 weld. Studies of the Gundremmingen reactor vessel, representing a joint USA-FRG-UK undertaking revealed an anomaly in strong vs. weak test orientation radiation sensitivity. (orig./HP)

  13. Advance of investigation of irradiation embrittlement mechanism of nuclear reactor pressure vessel steels. History and future of irradiation embrittlement researches

    Ishino, Shiori

    2007-01-01

    The nuclear reactor pressure vessel is the most important component of LWR plants required to be safe. This paper describes contents of the title consisting of four chapters. The first chapter states the general theory of irradiation effects, irradiation embrittlement and decreasing of toughness, and some kinds of pressure vessel steels. The second chapter explains history of irradiation embrittlement investigations and the advance of research methods for experiments and calculation. The third chapter contains information of inner structure of irradiated materials and development of prediction equations, recent information of embrittlement mechanism and mechanism guided prediction method, USA model and Central Research Institute of Electric Power Industry (CRIEPI) model. The fourth chapter states recent problems from viewpoints of experimental and analytical approaches. Comparison of standards of LWR pressure vessel steels between Japan and USA, relation between the density of number of cluster and the copper content, effect of flux on clustering of copper atoms, and CRIEPI's way of approaching the prediction method are illustrated. (S.Y.)

  14. Modeling irradiation embrittlement in reactor pressure vessel steels

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  15. Heavy-Section Steel Irradiation Program: Embrittlement issues

    Corwin, W.R.

    1991-01-01

    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 and the potential for major contamination releases. It is imperative to understand and predict the capabilities and limitations of its integrity. It is particularly 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 which would result in RPV failure. The Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) 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. Results from HSSI studies provide information needed to aid in resolving major regulatory issues facing the USNRC which 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 (LUS) welds. Taken together the results of these studies also provide guidance and bases for evaluating both the aging behavior and the potential for plant life extension of light-water RPVs. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. Embrittlement modeling studies have shown that the time or dose required for the point defect concentrations, which ultimately contribute to irradiation embrittlement, to reach their steady state values can be comparable to the component lifetime or to the duration of an irradiation experiment

  16. Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

    Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

    2007-07-01

    Full text of publication follows: Neutron irradiation of 9-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects that cause an increase in yield stress and ultimate tensile strength. This irradiation hardening causes embrittlement, which is observed in Charpy impact and toughness tests as an increase in ductile-brittle transition temperature (DBTT). Based on observations that show little change in strength in these steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above this irradiation-hardening temperature regime. In a recent study of F82H steel irradiated at 300, 380, and 500 deg. C, irradiation hardening-an increase in yield stress-was observed in tensile specimens irradiated at the two lower temperatures, but no change was observed for the specimens irradiated at 500 deg. C. As expected, an increase in DBTT occurred for the Charpy specimens irradiated at 300 and 380 deg. C. However, there was an unexpected increase in the DBTT of the specimens irradiated at 500 deg. C. The observed embrittlement was attributed to the irradiation-accelerated precipitation of Laves phase. This conclusion was based on results from a detailed thermal aging study of F82H, in which tensile and Charpy specimens were aged at 500, 550, 600, and 650 deg. C to 30,000 h. These studies indicated that there was a decrease in yield stress at the two highest temperatures and essentially no change at the two lowest temperatures. Despite the strength decrease or no change, the DBTT increased for Charpy specimens irradiated at all four temperatures. Precipitates were extracted from thermally aged specimens, and the amount of precipitate was correlated with the increase in transition temperature. Laves phase was identified in the extracted precipitates by X-ray diffraction. Earlier studies on conventional elevated-temperature steels also showed embrittlement effects above the irradiation-hardening temperature

  17. Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

    Najzer, M.

    1982-01-01

    Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

  18. Irradiation embrittlement and optimisation of annealing

    1993-01-01

    This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects

  19. Irradiation embrittlement and optimisation of annealing

    NONE

    1994-12-31

    This conference is composed of 30 papers grouped in 6 sessions related to the following themes: neutron irradiation effects in pressure vessel steels and weldments used in PWR, WWER and BWR nuclear plants; results from surveillance programmes (irradiation induced damage and annealing processes); studies on the influence of variations in irradiation conditions and mechanisms, and modelling; mitigation of irradiation effects, especially through thermal annealing; mechanical test procedures and specimen size effects.

  20. Review of recent studies on neutron irradiation embrittlement in light water reactor pressure vessel steels

    Sudo, Akira; Miyazono, Shohachiro

    1983-06-01

    Recent studies in foreign countries (USA, France, FRG and UK) on neutron irradiation embrittlement have been reviewed. These studies are classified into four areas, such as 1) effect of chemical composition on irradiation embrittlement sensitivity, 2) postirradiation heat treatment for embrittlement relief, 3) fracture toughness evaluation of irradiated materials based on fracture mechanics analysis, and 4) effect of irradiation on fatigue crack propagation behavior. The first area mainly includes the studies related to the effects of copper, phosphorus impurities and nickel alloying and synergistic effect of these components, and furthermore, evaluation of Regulatory Guide 1.99 Rev.l. Studies in the second area show the effects of annealing condition (temperature and time) and metallugical condition on embrittlement relief, and evaluation of periodic annealing in the period of irradiation as a promising method for embrittlement control. Studies in the third area show the correlation between fracture toughness and Cv notch ductility changes with neutron irradiation, and J-R curves of irradiated materials based on the elasto-plastic fracture mechanics. In the forth area, most of studies are investigated in air condition but a few studies in reactor-grade water at high temperature and pressure. (author)

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

    Marini, B., E-mail: bernard.marini@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France); Averty, X. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SEMI (now DEN/DANS/DM2S/SEMT), F-91191 Gif-sur Yvette (France); Wident, P.; Forget, P.; Barcelo, F. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France)

    2015-10-15

    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.

  2. Development of neutron irradiation embrittlement correlation of reactor pressure vessel materials of light water reactors

    Soneda, Naoki; Dohi, Kenji; Nomoto, Akiyoshi; Nishida, Kenji; Ishino, Shiori

    2007-01-01

    A large amount of surveillance data of the RPV embrittlement of the Japanese light water reactors have been compiled since the current Japanese embrittlement correlation has been issued in 1991. Understanding on the mechanisms of the embrittlement has also been greatly improved based on both experimental and theoretical studies. CRIEPI and the Japanese electric power utilities have started research project to develop a new embrittlement correlation method, where extensive study of the microstructural analyses of the surveillance specimens irradiated in the Japanese commercial reactors has been conducted. The new findings obtained from the experimental study are that the formation of solute-atom clusters with little or no copper is responsible for the embrittlement in low-copper materials, and that the flux effect exists especially in high-copper materials and this is supported by the difference in the microstructure of the high-copper materials irradiated at different fluxes. Based on these new findings, a new embrittlement correlation method is formulated using rate equations. The new methods has higher prediction capability than the current Japanese embrittlement correlation in terms of smaller standard deviation as well as smaller mean value of the prediction error. (author)

  3. Irradiation embrittlement of reactor vessel steels

    Bros, J.

    2000-01-01

    From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

  4. Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

    Burke, M.G.; Freyer, P.D.; Mager, T.R.

    1993-01-01

    In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ''precipitation-type'' and a ''damage-type'' component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs

  5. Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

    Burke, M G; Freyer, P D; Mager, T R

    1994-12-31

    In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ``precipitation-type`` and a ``damage-type`` component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs.

  6. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  7. Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

    Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

    2014-01-01

    The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

  8. Beryllium irradiation embrittlement test programme. Material and specimen specification, manufacture and qualification

    Harries, D.R.; Dalle Donne, M.

    1996-06-01

    The report presents the specification, manufacture and qualification of the beryllium specimens to be irradiated in the BR2 reactor in Mol to investigate the effect of the neutron irradiation on the embrittlement as a function of temperature and beryllium oxide content. This work was been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Union within the European Fusion Technology Program. (orig.)

  9. The modelling of irradiation embrittlement in submerged-arc welds

    Bolton, C.J.; Buswell, J.T.; Jones, R.B.; Moskovic, R.; Priest, R.H.

    1996-01-01

    Until very recently, the irradiation embrittlement behavior of submerged-arc welds has been interpreted in terms of two mechanisms, namely a matrix damage component and an additional component due to the irradiation-enhanced production of copper-rich precipitates. However, some of the weld specimens from a recent accelerated re-irradiation experiment have shown high Charpy shifts which exceeded the values expected from the measured shift in yield stress. Microstructural examination has revealed the occurrence of intergranular fracture (IGF) in these specimens, accompanied by grain boundary segregation of phosphorus. Theoretical models were developed to predict the parametric dependence of irradiation-enhanced phosphorus segregation on experimental variables. Using these parametric forms, along with the concept of a critical level of segregation for the onset of IGF instead of cleavage, a three mechanism trend curve has been developed. The form of this trend curve, taking into account IGF as well as matrix and copper embrittlement, is thus mechanistically based. The constants in the equation, however, are obtained by a statistical fit to the actual Charpy shift database

  10. Effect of chemical composition on irradiation embrittlement and annealing in Ni-Cr-Mo-V reactor pressure vessel steel

    Novosad, P [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    Results concerning copper and phosphorus influence on radiation-induced changes in the Ni-Cr-Mo-V steel mechanical properties, are presented. Correlations between different mechanical properties for steels with different chemical composition, are presented. A comparison of transition temperature shifts obtained for static and dynamic fracture toughness tests and Charpy impact tests, is discussed. Recovery of radiation hardening, measured by hardness test after isochronal annealing of steels with different compositions, is also shown. Copper content strongly affects irradiation-induced changes of mechanical properties, but phosphorus content in connection with variable copper content has only a small effect. (author). 4 refs., 4 figs., 4 tabs.

  11. R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels

    Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F.

    2000-01-01

    Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study thye evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (Author)

  12. Irradiation embrittlement of some 15Kh2MFA pressure vessel steels under varying neutron fluence rates

    Valo, M; Bars, B [Technical Research Centre of Finland, Espoo (Finland); Ahlstrand, A [Imatran Voima Oy (IVO), Helsinki (Finland)

    1994-12-31

    Irradiation sensitivity of two forging materials was measured with Charpy-V and fracture mechanic tests, and with different fluence, fluence rate and irradiation time values. Irradiation sensitivity of the materials was found to be less or equal to the current Russian standard, and appears to be well described by the fluence parameter only. A slight additional effect on embrittlement from a long term low fluence irradiation is noticed, but it stays within the total scatter band of data. 7 refs., 17 figs., 4 tabs.

  13. Modeling of irradiation embrittlement and annealing/recovery in pressure vessel steels

    Lott, R.G.; Freyer, P.D.

    1996-01-01

    The results of reactor pressure vessel (RPV) annealing studies are interpreted in light of the current understanding of radiation embrittlement phenomena in RPV steels. An extensive RPV irradiation embrittlement and annealing database has been compiled and the data reveal that the majority of annealing studies completed to date have employed test reactor irradiated weldments. Although test reactor and power reactor irradiations result in similar embrittlement trends, subtle differences between these two damage states can become important in the interpretation of annealing results. Microstructural studies of irradiated steels suggest that there are several different irradiation-induced microstructural features that contribute to embrittlement. The amount of annealing recovery and the post-anneal re-embrittlement behavior of a steel are determined by the annealing response of these microstructural defects. The active embrittlement mechanisms are determined largely by the irradiation temperature and the material composition. Interpretation and thorough understanding of annealing results require a model that considers the underlying physical mechanisms of embrittlement. This paper presents a framework for the construction of a physically based mechanistic model of irradiation embrittlement and annealing behavior

  14. Neutron irradiation embrittlement of reactor pressure vessel steels

    Steele, L.E.

    1975-01-01

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  15. Embrittling effects of residual elements on steels

    Brear, J.M.; King, B.L.

    1979-01-01

    In a review of work related to reheat cracking in nuclear pressure vessel steels, Dhooge et al referred to work of the authors on the relative embrittling parameter for SA533B steels. The poor agreement when these parameters were applied to creep ductility data for SA508 class 2 lead the reviewers to conclude that the relative importance of impurity elements is a function of base alloy composition. The authors briefly describe some of their more recent work which demonstrates that when various mechanical, and other, effects are taken into consideration, the relative effects of the principal residual elements are similar, despite differing base compositions, and that the embrittling parameters derived correlate well with the data for SA Class 2 steel. (U.K.)

  16. Severe embrittlement of neutron irradiated austenitic steels arising from high void swelling

    Neustroev, V.S. [FSUE ' SSC RF Research Institute of Atomic Reactors' , Dimitrovgrad (Russian Federation)], E-mail: neustroev@niiar.ru; Garner, F.A. [Pacific Northwest National Laboratory, Richland, WA (United States)

    2009-04-30

    Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components and introducing limitations on low temperature handling especially. It is shown that the degradation is actually a form of quasi-embrittlement arising from intense flow localization with high levels of localized ductility involving micropore coalescence and void-to-void cracking. Voids initially serve as hardening components whose effect is overwhelmed by the void-induced reduction in shear and Young's moduli at high swelling levels. Thus the alloy appears to soften even as the ductility plunges toward zero on a macroscopic level although a large amount of deformation occurs microscopically at the failure site. Thus the failure is better characterized as 'quasi-embrittlement' which is a suppression of uniform deformation. This case should be differentiated from that of real embrittlement which involves the complete suppression of the material's capability for plastic deformation.

  17. Relationship between irradiation hardening and embrittlement of pressure vessel steels

    Odette, G.R.; Lombrozo, P.M.; Wullaert, R.A.

    1984-01-01

    Based on a large body of test and power reactor data, empirical relationships between irradiation strengthening and embrittlement are derived. It is shown that the Charpy V-notch (C /SUB v/ ) 41-J indexed transition temperature increases and the upper-shelf energy decreases systematically with increases in the yield stress. The transition temperature shifts are related to two mechanisms: increases in the maximum temperature of elastic-cleavage fracture, and decreases in the slope of the C, energy versus test temperature curve associated with reductions in the upper-shelf energy. The cleavage shift contribution, which is usually dominant, can be predicted from the initial temperature of fracture at general yield and the change in ambient temperature static yield stress. In developing this simplified cleavage fracture model, it is shown that: (a) yield stress changes are independent of temperature and strain rate; (b) the increase in yield stress with decreasing temperature is independent of the strain rate, irradiation, and metallurgical state; and (c) the microcleavage fracture stress is independent of irradiation and temperature. A semi-empirical procedure for estimating the shift contribution due to upper-shelf energy decreases and the total temperature shift at 41 J, based on the observation of an approximately constant temperature interval of the transition regime, is proposed, along with a method for forecasting the entire irradiated C, curve

  18. A review of formulas for predicting irradiation embrittlement of reactors vessel materials

    Petrequin, P.

    1995-01-01

    Formulas developed in different countries for predicting irradiation embrittlement of reactors vessel materials are presented. Results of predictions were compared with different data sets, from surveillance programmes or studies in test reactors, with different residual elements contents. Figs

  19. Correlation methodology for predicting in-service irradiation embrittlement of reactor pressure vessels

    Odette, G.R.

    1980-01-01

    Irradiation embrittlement of reactor pressure steels is the consequence of altered microstructure due to both irradiation and time-at-temperature. Relatively poor characterisation of the initial microstructure and chemistry, and inaccurate dosimetry and temperature control, as well as failure properly to correlate these variables, have all contributed to a very large scatter in the experimental embrittlement data base. This has made improvement of the basic understanding of embrittlement very difficult. Therefore, it is necessary to develop a more realistic approach to utilising the data base. This is discussed, and proposals are made. (author)

  20. Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

    Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

  1. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    Takamizawa, Hisashi, E-mail: takamizawa.hisashi@jaea.go.jp; Itoh, Hiroto, E-mail: ito.hiroto@jaea.go.jp; Nishiyama, Yutaka, E-mail: nishiyama.yutaka93@jaea.go.jp

    2016-10-15

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

  2. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  3. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  4. The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

    Jones, R.B.; Buswell, J.T.

    1987-02-01

    An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 10 19 n.cm -2 (E>1.0MeV) at 250-290 0 C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu 3 P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

  5. Guidelines for prediction of irradiation embrittlement of operating WWER-440 reactor pressure vessels

    2005-06-01

    This TECDOC has been developed under an International Atomic Energy Agency Coordinated Research Project (CRP) entitled Evaluation of Radiation Damage of WWER Reactor Pressure Vessels (RPV) using Database on RPV Materials to develop the guidelines for prediction of radiation damage to WWER-440 PRVs. The WWER-440 RPV was designed by OKB Gidropress, Russian Federation, the general designer. Prediction of irradiation embrittlement of RPV materials is usually done in accordance with relevant codes and standards that are based on the large amounts of information from surveillance and research programmes. The existing Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than twenty years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. Nevertheless, it is still in use and generally consistent with new data. The present publication presents the analyses using all available data required for more precise prediction of radiation embrittlement of WWER-440 RPV materials. Based on the fact that it contains a large amount of data from surveillance programmes as well as research programmes, the IAEA International Database on RPV Materials (IDRPVM) is used for the detailed analysis of irradiation embrittlement of WWER RPV materials. Using IDRPVM, the guideline is developed for assessment of irradiation embrittlement of RPV ferritic materials as a result of degradation during operation. Two approaches, i.e. transition temperatures based on Charpy impact notch toughness, as well as based on static fracture toughness tests, are used in RPV integrity evaluation. The objectives of the TECDOC are the analysis of irradiation embrittlement data for WWER- 440 RPV materials using IDRPVM database, evaluation of predictive formulae depending on chemical composition of the material, neutron fluence, flux, and

  6. Microstructural design of PCA austenitic stainless steel for improved resistance to helium embrittlement under HFIR irradiation

    Maziasz, P.J.; Braski, D.N.

    1983-01-01

    Several variants of Prime Candidate Alloy (PCA) with different preirradiation thermal-mechanical treatments were irradiated in HFIR and were evaluated for embrittlement resistance via disk-bend tensile testing. Comparison tests were made on two heats of 20%-cold-worked type 316 stainless steel. None of the alloys were brittle after irradiation at 300 to 400 0 C to approx. 44 dpa and helium levels of 3000 to approx.3600 at. ppm. However, all were quite brittle after similar exposure at 600 0 C. Embrittlement varied with alloy and pretreatment for irradiation to 44 dpa at 500 0 C and to 22 dpa at 600 0 C. Better relative embrittlement resistance among PCA variants was found in alloys which contained prior grain boundary MC carbide particles that remained stable under irradiation

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

    Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation); National Research Nuclear University “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, Moscow 115409 (Russian Federation); Gurovich, B.A.; Bukina, Z.V.; Frolov, A.S.; Maltsev, D.A.; Krikun, E.V.; Zhurko, D.A.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

    2017-07-15

    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 (ΔT{sub K}) 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 ΔT{sub K} 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 ΔT{sub K} shift in the studied range of irradiation temperature and fluence. - Highlights: •Structural elements in RPV steel are studied at different irradiation temperatures. •Highest number density dislocation loops are

  8. Comparative study for the estimation of To shift due to irradiation embrittlement

    Lee, Jin Ho; Park, Youn won; Choi, Young Hwan; Kim, Seok Hun; Revka, Volodymyr

    2002-01-01

    Recently, an approach called the 'Master Curve' method was proposed which has opened a new means to acquire a directly measured material-specific fracture toughness curve. For the entire application of the Master Curve method, several technical issues should be solved. One of them is to utilize existing Charpy impact test data in the evaluation of a fracture transition temperature shift due to irradiation damage. In the U.S. and most Western countries, the Charpy impact test data have been used to estimate the irradiation effects on fracture toughness changes of RPV materials. For the determination of the irradiation shift the indexing energy level of 41 joule is used irrespective of the material yield strength. The Russian Code also requires the Charpy impact test data to determine the extent of radiation embrittlement. Unlike the U.S. Code, however, the Russian approach uses the indexing energy level varying according to the material strength. The objective of this study is to determine a method by which the reference transition temperature shift (ΔT o ) due to irradiation can be estimated. By comparing the irradiation shift estimated according to the U.S. procedure (ΔT 41J ) with that estimated according to the Russian procedure (ΔT F ), it was found that one-to-one relation exists between ΔT o and ΔT F

  9. Influence of helium embrittlement on post-irradiation creep rupture behaviour of austenitic and martensitic stainless steels

    Wassilew, C.

    1982-01-01

    The author has investigated the influence of helium embrittlement on the creep rupture properties of the austenitic stainless steels 1.4970 and 1.4962 and the martensitic stainless steel 1.4914 after irradiation in the BR-2 reactor in Mol, Belgium. The results show that austenitic steels react much more strongly to the embrittlement effect of the helium than do martensitic steels. The causes of the lower embrittlement tendency of the martensitic than of both austenitic stainless steels were analysed carefully. A new embrittlement model was developed on the basis of data derived from the creep rupture experiments, and reinforced by a simple metallographic investigation of the fracture zone and its immediate environment. This model pays specific attention to the role of the twin planes as the most efficient area of increased vacancy production, and takes into account the ability of the twin boundaries to transport these vacancies with reduced energy and low loss into the high-angle grain boundaries. (author)

  10. Severe Embrittlement of Neutron Irradiated Austenitic Steels Arising from High Void Swelling

    Neustroev, V.S.; Garner, F.

    2007-01-01

    Full text of publication follows: Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components. Similar loss of ductility is expected when swelling arises in fusion and light water reactor environments. Above 7-16% swelling there is complete loss of ductility, with the onset of ductility loss beginning at lower swelling in ring-pull tensile tests than for flat tensile specimens. For steels that develop extensive precipitation during irradiation, the critical swelling level is even lower. A model is presented to demonstrate the effect of voids acting alone to produce the embrittlement. Although voids are not very effective hardeners, they are very effective to generate stress concentrations between voids. The stress concentration ratio increases strongly when the void diameter exceeds ∼40% of the void-to-void separation distance. When the volume fraction of voids is rather high (about 16 % and higher), a geometric situation develops where it is possible to create an intense field of deformation glide planes residing at an angle of 45 deg. to the void-to-void axis. Significant localized flow then proceeds on these planes for specimen stress levels that are significantly lower than the yield stress. Voids also segregate nickel to their surfaces such that flow localization occurs in the low-nickel inter-void regions to produce strain-induced martensite, which is further accelerated by stress concentrations at the advancing crack tip, leading to catastrophic failure. (authors)

  11. Heavy-section steel irradiation program: Embrittlement issues

    Corwin, W.R.

    1991-01-01

    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 and the potential for major contamination releases. The RPV is one of only two major safety- related components 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 which would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) 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 including fracture toughness crack arrest toughness ductile tearing resistance Charpy V-notch impact energy, dropweight nil-ductility temperature and tensile properties. Models based on observations of radiation-induced microstructural changes using the field on microprobe and the high resolution transmission electron microscopy provide improved bases for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs

  12. Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

    Ghoneim, M.M.

    1987-05-01

    The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

  13. SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Metal Irradiation Embrittlement, annealing and Re-Embrittlement. Second Progress Report

    Van Walle, E.; Chaouadi, R.; Scibetta, M.; Lucon, E.; Weber, M.

    1999-07-01

    The report gives the actual status of the contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement. Results from the reference testing of unirradiated material as well as the results of the CHIVAS-7 experiment are discussed

  14. Irradiation Embrittlement Monitoring Programs of RPV's in the Slovak Republic NPP's

    Kupca, Ludovik

    2006-01-01

    Four types of surveillance programs were (are) realized in Slovak NPP's: 'Standard Surveillance Specimen Program' (SSSP) was finished in Jaslovske Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4, 'Extended Surveillance Specimen Program' (ESSP), was prepared for Jaslovske Bohunice NPP V-2 with aim to validate the SSSP results, For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program 'Modern Surveillance Specimen Program' (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life, For the Bohunice V-1 NPP was finished 'New Surveillance Specimen Program' (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years, New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV's austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: increasing of neutron fluence measurement accuracy, substantial improvement the irradiation temperature measurement, fixed orientation of samples to the centre of the reactor core, minimum differences of neutron dose for all the Charpy-V notch and COD specimens, the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV's unit-1 and 2, based on the fundamental postulate - to provide the irradiation embrittlement monitoring till the end of units operation. The 'New Surveillance Specimen Program' (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The

  15. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

    Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-05-01

    Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

  16. Embrittlement of a 17Cr ferritic steel irradiated in Phenix

    Allegraud, G.; Boutard, J.L.; Boyer, J.M.

    1987-01-01

    Charpy V and tensile tests have been performed with samples made of 17Cr ferritic steel irradiated in Phenix at temperatures between 390 and 540C up to a maximum dose of 83.3 dpaF. All over the temperature and dose ranges, irradiation leads to an increase of the ductile brittle transition temperature (DBTT). The DBTT and hardening are decreasing functions of the irradiation temperature. Fast neutron flux at 390C hardens the material more than a sole thermal ageing does

  17. Evaluation of the french test reactors irradiation embrittlement experiments

    Miannay, D.; Dussarte, D.; Soulat, P.

    1988-07-01

    The shifts of CV 41J energy index temperatures due to irradiation measured in France have been stored in a data bank and are analysed. According to a simple physically based model which is here-after verified, correlations are proposed for Base Metal (BM) and Weld Metal (WM). The achemical and phosphorus components of the chemical factor are equivalent. However, nickel and copper play a leading part in BM and WM respectively. The copper nickel interaction is not evident. These correlations are for cleavage fracture and not for intergranular fracture. This work is subject to revision and extension

  18. Temper embrittlement, irradiation induced phosphorus segregation and implications for post-irradiation annealing of reactor pressure vessels

    McElroy, R.J.; English, C.A.; Foreman, A.J.; Gage, G.; Hyde, J.M.; Ray, P.H.N.; Vatter, I.A.

    1999-01-01

    Three steels designated JPB, JPC and JPG from the IAEA Phase 3 Programme containing two copper and phosphorus levels were pre- and post-irradiation Charpy and hardness tested in the as-received (AR), 1200 C/0.5h heat treated (HT) and heat treated and 450 C/2000h aged (HTA) conditions. The HT condition was designed to simulate coarse grained heat-affected zones (HAZ's) and showed a marked sensitivity to thermal ageing in all three alloys. Embrittlement after thermal ageing was greater in the higher phosphorus alloys JPB and JPG. Charpy shifts due to thermal ageing of between 118 and 209 C were observed and accompanied by pronounced intergranular fracture, due to phosphorus segregation. The irradiation embrittlement response was complex. The low copper alloys, JPC and JPB, in the HT and HTA condition exhibited significant irradiation induced Charpy shift but very low or even negative hardness changes indicating non-hardening embrittlement. The higher copper alloy, JPG, also exhibited irradiation hardening in line with its copper content. Fractographic and microchemical studies indicated irradiation induced phosphorus segregation and a transition from cleavage to intergranular failure at grain boundary phosphorus concentrations above a critical level. The enhanced grain boundary phosphorus level increased with dose in agreement with a kinetic segregation model developed at Harwell. The relevance of the thermal ageing studies to RPV Annealing for Plant-Life Extension was identified early in the program. It is of concern that annealing of RPV's has been performed, or is proposed, at temperatures in the range 425--475 C for periods of about 1 week (168h). Much attention has been given to the use of in-situ hardness measurements and machining miniature Charpy and tensile specimens from belt-line plate and weld materials. However, HAZ's, often containing higher phosphorus levels than the present materials, have largely been ignored. A post-irradiation annealing (PIA

  19. Neutron-irradiation + helium hardening and embrittlement modeling of 9% Cr-steels in an engineering perspective (HELENA)

    Chaouadi, Rachid

    2008-07-01

    This report provides a physically-based engineering model to estimate the radiation hardening of 9%Cr-steels under both displacement damage (dpa) and helium. The model is essentially based on the dispersed barrier hardening theory and the dynamic re-solution of helium under displacement cascades. However, a number of assumptions and simplifications were considered to obtain a simple description of irradiation hardening and embrittlement primarily relying on the available experimental data. As a result, two components were basically identified, the dpa component that can be associated with black dots and small loops and the He-component accounting for helium bubbles. The dpa component is strongly dependent on the irradiation temperature and its dependence law was based on a first-order annealing kinetics. The damage accumulation law was also modified to take saturation into account. Finally, the global kinetics of the damage accumulation kept defined, its amplitude is fitted to one experimental condition. The model was rationalized on an experimental database that mainly consists of {proportional_to}9%Cr-steels irradiated in the technologically important temperature range of 50 to 600 C up do 50 dpa and with a He-content up to {proportional_to}5000 appm, including neutron and proton irradiation as well as implantation. The test temperature effect is taken into account through a normalization procedure based on the change of the Young's modulus and the anelastic deformation that occurs at high temperature. Finally, the hardening-to-embrittlement correlation is obtained using the load diagram approach. Despite the large experimental scatter, inherent to the variety of the materials and irradiation as well as testing conditions, the obtained results are very promising. Improvement of the model performance is still possible by including He-hardening saturation and high temperature softening but unfortunately, at this stage, a number of conflicting experimental data

  20. Neutron-irradiation + helium hardening and embrittlement modeling of 9% Cr-steels in an engineering perspective (HELENA)

    Chaouadi, Rachid

    2008-01-01

    This report provides a physically-based engineering model to estimate the radiation hardening of 9%Cr-steels under both displacement damage (dpa) and helium. The model is essentially based on the dispersed barrier hardening theory and the dynamic re-solution of helium under displacement cascades. However, a number of assumptions and simplifications were considered to obtain a simple description of irradiation hardening and embrittlement primarily relying on the available experimental data. As a result, two components were basically identified, the dpa component that can be associated with black dots and small loops and the He-component accounting for helium bubbles. The dpa component is strongly dependent on the irradiation temperature and its dependence law was based on a first-order annealing kinetics. The damage accumulation law was also modified to take saturation into account. Finally, the global kinetics of the damage accumulation kept defined, its amplitude is fitted to one experimental condition. The model was rationalized on an experimental database that mainly consists of ∝9%Cr-steels irradiated in the technologically important temperature range of 50 to 600 C up do 50 dpa and with a He-content up to ∝5000 appm, including neutron and proton irradiation as well as implantation. The test temperature effect is taken into account through a normalization procedure based on the change of the Young's modulus and the anelastic deformation that occurs at high temperature. Finally, the hardening-to-embrittlement correlation is obtained using the load diagram approach. Despite the large experimental scatter, inherent to the variety of the materials and irradiation as well as testing conditions, the obtained results are very promising. Improvement of the model performance is still possible by including He-hardening saturation and high temperature softening but unfortunately, at this stage, a number of conflicting experimental data reported in literature should

  1. Results from Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants - Irradiation Embrittlement of RPV Steels -

    Abe, Hiroaki; Onizawa, Kunio; Katsuyama, Jinya; Murakami, Kenta; Iwai, Takeo; Iwata, Tadao; Katano, Yoshio; Sekimura, Naoto; Nagai, Yasuyoshi; Toyama, Takeshi; Tamura, Satoshi

    2012-01-01

    As one of the NISA Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants, we have performed research on the irradiation embrittlement of reactor pressure vessel (RPV) steels, especially focusing on irradiation embrittlement on heat affected zone (HAZ) and on applications of ion beams to deduce fundamental insights irradiation-induced embrittlement. The results obtained from the project are summarized as follows. In order to obtain the technical basis to judge the necessity of surveillance specimens from HAZ, the neutron irradiation program was performed at JRR-3, JAEA. The samples were carefully designed based on the insights from finite element analysis, metallography, 3D atom probe and positron annihilation methods, and were fabricated so as to simulate both heat treatment history and microstructure for typical HAZ from as-fabricated RPV steels which also have variation of impurity levels. The fracture toughness of the unirradiated HAZ specimens was equivalent to or better than that of base metals. Irradiation embrittlement and hardening were roughly identical to those of base metals, while some of the fine-grained HAZ microstructure was susceptible to it. The probabilistic fracture mechanics analysis was applied to the structural integrity assessment taking into account the heterogeneous microstructure as well as susceptibility for irradiation embrittlement of each HAZ microstructure under the variation of welding parameter and PTS condition. It was shown that crack propagation at the fine-grained HAZ, but the discontinuous distribution of the microstructure retards the further propagation. For the precise correlation of irradiation embrittlement of RPV steels for the long term operations, accumulations of high-dose data are required. Ion beam irradiation is one of the solutions for the regime and for mechanism-based descriptions. Another interest of ours was to describe irradiation hardening and embrittlement in terms of

  2. A study of the mechanical property changes of irradiation embrittled pressure vessel steels and their response to annealing treatments

    Tipping, P.; Waeber, W.B.; Mercier, O.

    1991-01-01

    Isochronal and isothermal heat treatments have been used to study the recovery of hardness of a neutron irradiated pressure vessel steel forging for the purposes of planning and realizing IAR (Irradiated-Annealed-Reirradiated) experiments. Charpy V notch tests have been performed to assess the toughness of the material irradiated to various fluences up to a maximum of 5 x 10 19 n/cm 2 , E>1 MeV at 290 o C with and without an intermediate annealing treatment at 450 o C x 168 h. The effect of the intermediate annealing was evident. The recovery of the upper shelf energies was strongly enhanced by a thermal ageing effect due to the annealing treatment for all fluence levels investigated compared to the irradiated condition. The transition temperature shifts exhibited a less straightforward behaviour due to the mentioned ageing effect which opposed the recovery process for this property leading to a net shift increase at lower and to a net recovery benefit at higher fluence levels. A phenomenological model description for the IAR embrittlement-recovery path is suggested. For this material and these irradiation conditions a plant life extension (PLEX) may be brought about if a specific annealing treatment is applied at a fluence level that is half the anticipated target fluence F for PLEX. In this case it was found that F>1.6 x 10 19 n/cm 2 . (author)

  3. The irradiation embrittlement of two pressure vessel steels -Contribution of local approach

    Soulat, P; Marini, B [CEA Centre d` Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Miannay, D; Horowitz, H [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Schill, R [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

    1994-12-31

    Within the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses``, the French participation has been focused on the contribution of the local approach to the determination of the sensitivity to radiation embrittlement of two different pressure vessel steels: a low sensitive French forging steel (FFA) and a high sensitive ``monitor`` Japanese plate steel (JRQ) were irradiated to a fluence of 3.10{sup 19} n/cm{sup 2} at 290 C. The irradiation embrittlement of the two steels measured by the shift of Charpy V transition curves is in good agreement with the estimated shifts given by theoretical prediction. The fracture toughness properties were examined at low temperature with brittle fracture, and at service temperature (290 C), with ductile tearing. The values of K{sub 1C} or K{sub JC} for the brittle fracture and J{sub 1C} for the ductile fracture are compared to predictions established using the local approach of cleavage fracture (Weibull analysis) and the critical rate of void growth respectively. 8 refs., 14 figs., 10 tabs.

  4. Hardening and embrittlement mechanisms of reduced activation ferritic/martensitic steels irradiated at 573 K

    Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Hashimoto, N. [Hokkaido Univ., Materials Science and Engineering Div., Graduate School of Engineering, Sapporo (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

    2007-07-01

    Full text of publication follows: It has been reported that reduced-activation ferritic/martensitic steels (RAFMs), such as F82H, ORNL9Cr-2WVTa, and JLF-1, showed a variety of changes in ductile-brittle transition temperature and yield stress after irradiation at 573 K up to 5 dpa, and those differences could not be interpreted solely by the difference of dislocation microstructure induced by irradiation. To investigate the impact of other microstructural feature, i.e. precipitates, the precipitation behavior of F82H, ORNL 9Cr-2WVTa, and JLF-1 was examined. It was revealed that irradiation-induced precipitation and amorphization of precipitates partly occurred and caused the different precipitation on block, packet and prior austenitic grain boundaries. In addition to these phenomena, irradiation-induced nano-size precipitates were also observed in the matrix. It was also revealed that the chemical compositions of precipitates approached the calculated thermal equilibrium state of M{sub 23}C{sub 6} at an irradiation temperature of 573 K. The calculation also suggests the presence of Laves phase at 573 K, which is usually not observed at this temperature, but the ion irradiation on aged F82H with Laves phase suggests that Laves phase becomes amorphous and could not be stable under irradiation at 573 K. This observation indicates the possibility that the irradiation-induced nano-size precipitation could be the consequence of the conflict between precipitation and amorphization of Laves phase. Over all, these observations suggests that the variety of embrittlement and hardening of RAFMs observed at 573 K irradiation up to 5 dpa might be the consequence of the transition phenomena that occur as the microstructure approaches thermal equilibrium during irradiation at 573 K. (authors)

  5. Effect of lead factors on the embrittlement of RPV SA-508 cl 3 steel

    Kempf, Rodolfo, E-mail: kempf@cnea.gov.ar [CNEA, Unidad Actividad Combustibles Nucleares, División Caracterización, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina); Troiani, Horacio, E-mail: troiani@cab.cnea.gov.ar [Centro Atómico Bariloche (CNEA) e Instituto Balseiro (UNCU), CONICET, Av. Bustillo 9500, CP 8400, Rio Negro (Argentina); Fortis, Ana Maria, E-mail: fortis@cnea.gov.ar [CNEA, Departamento Estructura y Comportamiento, UNSAM, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina)

    2013-03-15

    This paper presents a project to study the effect of lead factors on the mechanical behaviour of the SA-508 type 3 Reactor Pressure Vessel (RPV) steel used in the reactor under construction Atucha II in Argentina. Charpy-V notch specimens of this steel were irradiated at the RA1 experimental reactor at a temperature of 275 °C with two lead factors (186 and 93). The neutron flux was 3.71 × 10{sup 15} n m{sup −2} s{sup −1} and 1.85 × 10{sup 15} n m{sup −2} s{sup −1} (E > 1 MeV) respectively. In both cases, the fluence was 6.6 × 10{sup 21} n m{sup −2}, which is equivalent to that received by the PHWR Atucha II RPV in 10 years of full power irradiation. The results of Charpy tests revealed significant embrittlement both in the ΔT = 14 °C and ΔT = 21 °C shifts of the ductile–brittle transition temperatures (DBTT) and in the reduction of the maximum energy absorbed. This result shows that the shift of the DBTT with a lead factor of 93 is larger than that obtained with a lead factor of 186. Then, the results of irradiation in experimental reactors (MTR) with high lead factors may not be conservative with respect to the actual RPV embrittlement.

  6. Irradiation embrittlement and mitigation. V. 1. Working material. Proceedings of a specialists meeting held in Espoo, Finland 23-26 October 1995

    1995-01-01

    The purpose of the meeting was to provide an international forum for discussion on recent results in research and utility experience on radiation damage and its surveillance, annealing and re-embrittlement of PWR, WWER and BWR reactor pressure vessel materials. The scope included: mechanism of radiation damage; effects of operating parameters (flux, temperature, time, etc.); results from surveillance programmes and their analysis; fracture mechanics testing and evaluation; annealing and optimization of the process; re-embrittlement after annealing. Presentations were aimed at better understanding of radiation damage, annealing and re-irradiation behaviour of reactor pressure vessels materials, at providing guidance and recommendations for optimization of annealing and surveillance programmes and directions for further investigations. Refs, figs and tabs

  7. Irradiation embrittlement and mitigation. V. 1. Working material. Proceedings of a specialists meeting held in Espoo, Finland 23-26 October 1995

    NONE

    1996-12-31

    The purpose of the meeting was to provide an international forum for discussion on recent results in research and utility experience on radiation damage and its surveillance, annealing and re-embrittlement of PWR, WWER and BWR reactor pressure vessel materials. The scope included: mechanism of radiation damage; effects of operating parameters (flux, temperature, time, etc.); results from surveillance programmes and their analysis; fracture mechanics testing and evaluation; annealing and optimization of the process; re-embrittlement after annealing; Presentations were aimed at better understanding of radiation damage, annealing and re-irradiation behaviour of reactor pressure vessels materials, at providing guidance and recommendations for optimization of annealing and surveillance programmes and directions for further investigations. Refs, figs and tabs.

  8. Effect of heat treatments on the hydrogen embrittlement ...

    pipe steel in as received (controlled rolled), normalized, and quenched and tempered conditions. The resistance to hydrogen embrittlement was found in the order of controlled rolled > quenched and tempered > normalized. The fracture mode ...

  9. Hydrogen Embrittlement

    Woods, Stephen; Lee, Jonathan A.

    2016-01-01

    Hydrogen embrittlement (HE) is a process resulting in a decrease in the fracture toughness or ductility of a metal due to the presence of atomic hydrogen. In addition to pure hydrogen gas as a direct source for the absorption of atomic hydrogen, the damaging effect can manifest itself from other hydrogen-containing gas species such as hydrogen sulfide (H2S), hydrogen chloride (HCl), and hydrogen bromide (HBr) environments. It has been known that H2S environment may result in a much more severe condition of embrittlement than pure hydrogen gas (H2) for certain types of alloys at similar conditions of stress and gas pressure. The reduction of fracture loads can occur at levels well below the yield strength of the material. Hydrogen embrittlement is usually manifest in terms of singular sharp cracks, in contrast to the extensive branching observed for stress corrosion cracking. The initial crack openings and the local deformation associated with crack propagation may be so small that they are difficult to detect except in special nondestructive examinations. Cracks due to HE can grow rapidly with little macroscopic evidence of mechanical deformation in materials that are normally quite ductile. This Technical Memorandum presents a comprehensive review of experimental data for the effects of gaseous Hydrogen Environment Embrittlement (HEE) for several types of metallic materials. Common material screening methods are used to rate the hydrogen degradation of mechanical properties that occur while the material is under an applied stress and exposed to gaseous hydrogen as compared to air or helium, under slow strain rates (SSR) testing. Due to the simplicity and accelerated nature of these tests, the results expressed in terms of HEE index are not intended to necessarily represent true hydrogen service environment for long-term exposure, but rather to provide a practical approach for material screening, which is a useful concept to qualitatively evaluate the severity of

  10. Modeling copper precipitation hardening and embrittlement in a dilute Fe-0.3at.%Cu alloy under neutron irradiation

    Bai, Xian-Ming; Ke, Huibin; Zhang, Yongfeng; Spencer, Benjamin W.

    2017-11-01

    Neutron irradiation in light water reactors can induce precipitation of nanometer sized Cu clusters in reactor pressure vessel steels. The Cu precipitates impede dislocation gliding, leading to an increase in yield strength (hardening) and an upward shift of ductile-to-brittle transition temperature (embrittlement). In this work, cluster dynamics modeling is used to model the entire Cu precipitation process (nucleation, growth, and coarsening) in a Fe-0.3at.%Cu alloy under neutron irradiation at 300°C based on the homogenous nucleation mechanism. The evolution of the Cu cluster number density and mean radius predicted by the modeling agrees well with experimental data reported in literature for the same alloy under the same irradiation conditions. The predicted precipitation kinetics is used as input for a dispersed barrier hardening model to correlate the microstructural evolution with the radiation hardening and embrittlement in this alloy. The predicted radiation hardening agrees well with the mechanical test results in the literature. Limitations of the model and areas for future improvement are also discussed in this work.

  11. Reactor pressure vessel embrittlement: Insights from neural network modelling

    Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

    2018-04-01

    Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

  12. Modeling of cavity swelling-induced embrittlement in irradiated austenitic stainless steels

    Han, X.

    2012-01-01

    During long-time neutron irradiation occurred in Pressurized Water Reactors (PWRs), significant changes of the mechanical behavior of materials used in reactor core internals (made of 300 series austenitic stainless steels) are observed, including irradiation induced hardening and softening, loss of ductility and toughness. So far, much effect has been made to identify radiation effects on material microstructure evolution (dislocations, Frank loops, cavities, segregation, etc.). The irradiation-induced cavity swelling, considered as a potential factor limiting the reactor lifetime, could change the mechanical properties of materials (plasticity, toughness, etc.), even lead to a structure distortion because of the dimensional modifications between different components. The principal aim of the present PhD work is to study qualitatively the influence of cavity swelling on the mechanical behaviors of irradiated materials. A micromechanical constitutive model based on dislocation and irradiation defect (Frank loops) density evolution has been developed and implemented into ZeBuLoN and Cast3M finite element codes to adapt the large deformation framework. 3D FE analysis is performed to compute the mechanical properties of a polycrystalline aggregate. Furthermore, homogenization technique is applied to develop a Gurson-type model. Unit cell simulations are used to study the mechanical behavior of porous single crystals, by accounting for various effects of stress triaxiality, of void volume fraction and of crystallographic orientation, in order to study void effect on the irradiated material plasticity and roughness at polycrystalline scale. (author) [fr

  13. Effect of trapping and temperature on the hydrogen embrittlement susceptibility of alloy 718

    Galliano, Florian; Andrieu, Eric; Blanc, Christine; Cloue, Jean-Marc; Connetable, Damien; Odemer, Gregory, E-mail: gregory.odemer@ensiacet.fr

    2014-08-12

    Ni-based alloy 718 is widely used to manufacture structural components in the aeronautic and nuclear industries. Numerous studies have shown that alloy 718 may be sensitive to hydrogen embrittlement. In the present study, the susceptibilities of three distinct metallurgical states of alloy 718 to hydrogen embrittlement were investigated to identify both the effect of hydrogen trapping on hydrogen embrittlement and the role of temperature in the hydrogen-trapping mechanism. Cathodic charging in a molten salt bath was used to saturate the different hydrogen traps of each metallurgical state. Tensile tests at different temperatures and different strain rates were carried out to study the effect of hydrogen on mechanical properties and failure modes, in combination with hydrogen content measurements. The results demonstrated that Ni-based superalloy 718 was strongly susceptible to hydrogen embrittlement between 25 °C and 300 °C, and highlighted the dominant roles played by the hydrogen solubility and the hydrogen trapping on mechanical behavior and fracture modes.

  14. Effect of hydrogen and oxygen content on the embrittlement of Zr alloys

    Griger, A.; Hozer, Z.; Matus, L.; Vasaros, L.; Horvath, M.

    2001-01-01

    An experimental study is carried out in the KFKI Atomic Energy Research Institute in order to clear up the role of oxidation and hydrogen uptake in the embrittlement process. Russian E110 type Zr1%Nb and Zircaloy-4 claddings are used as test materials. The differences between the properties of two alloys are examined. The sample preparation covered the following cases: oxidation in Ar+O 2 atmosphere; hydrogen uptake of as received and pre-oxidised samples (in Ar+O 2 atmosphere); oxidation in steam. The oxidation in Ar+O 2 and the subsequent hydrogen uptake procedure make possible the production of samples with well-characterized hydrogen and oxygen content. Corrosion treated ring samples of 8 mm height are examined in ring compression tests. The force-deformation curves are recorded and the crushing force and deformation are determined. The relative deformation is used for the characterisation of embrittlement level. The results of experiments provide detailed information about the effect of hydrogen and oxygen content on the embrittlement of zirconium alloys. The conclusions are: 1) hydrogen seems to play a more important role in the embrittlement of zirconium alloys than oxygen; 2) the Zircaloy-4 alloy becomes brittle at lower hydrogen content than the Zr1%Nb; 3) under steam oxidation conditions the Zr1%Nb alloy takes up much more hydrogen and becomes more brittle than the Zircaloy-4

  15. Modeling of helium effects in metals: High temperature embrittlement

    Trinkaus, H.

    1985-01-01

    The effects of helium on swelling, creep rupture and fatigue properties of fusion reactor materials subjected to (n,α)-reactions and/or direct α-injection, are controlled by bubble formation. The understanding of such effects requires therefore the modeling of (1) diffusional reactions of He atoms with other defects; (2) nucleation and growth of He bubbles; (3) transformation of such bubbles into cavities under continuous He generation and irradiation or creep stress. The present paper is focussed on the modeling of the (coupled) high temperature bubble nucleation and growth processes within and on grain boundaries. Two limiting cases are considered: di-atomic nucleation described by the simplest possible sets of rate equations, and multi-atomic nucleation described by classical nucleation theory. Scaling laws are derived which characterize the dependence of the bubble densities upon time (He-dose), He generation rate and temperature. Comparison with experimental data of AISI 316 SS α-implanted at temperatures around 1000 K indicates bubble nucleation of the multi-atomic type. The nucleation and growth models are applied to creep tests performed during α-implantation suggesting that in these cases gas driven bubble growth is the life time controlling mechanism. The narrow (creep stress/He generation rate) range of this mechanism in a mechanism map constructed from these tests indicates that in many reactor situations the time to rupture is probably controlled by stress driven cavity growth rather than by gas driven bubble growth. (orig.)

  16. Surveillance as a complement to irradiation embrittlement studies: Status and needs

    Steele, L.E.

    1977-01-01

    The history of the study of radiation embrittlement of reactor pressure vessel steels has gone through three stages in the USA. 1) A scientific curiosity. 2) Empirical or laboratory evaluation of typical steels, and 3) Integration of the scientific and empirical to advance status and evolve standard techniques. The current stage is one in which surveillance data compliments the laboratory studies which characterized Stage 3. The early USA surveillance programs were generally analyzed by the same people who were the primary laboratory investigators. An effort must be made to continue this type of collaboration as a useful two-way learning procedure though it will become more and more difficult as nuclear power is broadly commercialized. The current status of both types of USA programs will be presented to encourage the most advantageous use of data from both sources. At this time about 25 USA nuclear power reactors have operated long enough to have provided initial surveillance or dosimetry results. An effort will be made to summarize the general status of these in order to: 1) Provide complimentary data to laboratory studies. 2) Assess directions in handling the problems of radiation embrittlement. 3) Note lessons learned for improving surveillance efforts in the future. 4) Identify possible research tasks for the future to support in-service surveillance and other measures. 5) Justify facts advancing surveillance requirements to status of national codes and standards. 6) Justify facts requiring changes in current national codes and standards. A plan will be presented along with an introduction of each member of the USA delegation for systematic presentation of the status of reactor vessel surveillance in the USA. (author)

  17. Blistering and hydride embrittlement

    Louthan, M.R. Jr.

    1975-01-01

    The effects of hydrogen on the mechanical properties of metals have been categorized into several groups. Two of the groups, hydrogen blistering and hydride embrittlement, are reasonably well understood, and problems relating to their occurrence may be avoided if that understanding is used as a basis for selecting alloys for hydrogen service. Blistering and hydride embrittlement are described along with several techniques of materials selection and used to minimize their adverse effects. (U.S.)

  18. Evaluation on the Effect of Composition on Radiation Hardening and Embrittlement in Model FeCrAl Alloys

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Edmondson, Philip [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Xunxiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Littrell, Kenneth C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    networks acting as defect sinks, resulting in variations in the observed microstructures after irradiation. Dose trends were also observed, with increasing radiation dose promoting changes in the size and number density of the Cr-rich α' precipitates. Based on the microstructural analysis, performed tensile testing, and prior knowledge from FeCr literature it was hypothesized that the formation of the Cr-rich α' precipitates could lead to significant radiation-induced embrittlement in the alloys, and this could be composition dependent, a result which would mirror the trends observed for radiation-induced hardening. Due to the limited database on embrittlement in the FeCrAl alloy class after irradiation, a series of radiation experiments have been implemented. The overarching point of view within this report is the radiation tolerance of FeCrAl is complex, with many mechanisms and factors to be considered at once. Further development of the FeCrAl alloy class for enhanced accident tolerant applications requires detailed, single (or at least limited) variable experiments to fully comprehend and predict the performance of this alloy in LWRs. This report has been submitted as fulfillment of milestone M2FT-15OR0202321 titled, Summary report on the effect of composition on the irradiation embrittlement of Gen 1 ATF FeCrAl for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

  19. The metrological problems of irradiation embrittlement of reactor pressure vessel steel

    Vodenicharov, S.; Kamenova, Ts.

    1993-01-01

    Neutron irradiation of reactor pressure vessel steels increases the T k -values of transition temperature from ductile to brittle fracture. This effect is very important in emergency situations, when the water cooling injection in the reactor results in high thermal gradients. In such cases there is a risk from the appearance of a brittle fracture with catastrophic crack propagation speed at relatively low stresses. That is why the T k -value determination is very important for the safe operation of the reactor systems. Some advanced experimental methods for T k -testing and control have been discussed in the present article and the standards of different countries have been compared. The methods applying subsize specimens and welding-restored specimens have been reviewed. (author)

  20. Cooperation modes of the radiation embrittlement

    Voevodin, V.N.; Laptev, I.N.; Neklyudov, I.M.; Ozhigov, L.S.; Bryk, V.V.; Parkhomenko, A.A.

    2012-01-01

    According to the results of experimental and theoretical studies of the structures and properties of irradiated deformed materials with different crystalline structure, the effect of irradiation on mechanisms of radiation embrittlement on all structure levels (from atomic to macrolevel) has been shown. The effects of structural localization, collectivization, long range effects, rotation modes development are described. It was shown that these effects are closely interrelated; they characterized the deformed irradiation material as open dissipative system subjected to the laws of such scientific approach as synergetic.

  1. Assessment of the French and US embrittlement trend curves applied to RPV materials irradiated in the BR2 materials test reactor

    Chaouadi, R.; Gerard, R.; Boagaerts, A.S.

    2011-01-01

    The irradiation embrittlement of reactor pressure vessels (RPVs) in monitored through the surveillance programs associated with predictive formulas, the so-called embrittlement trend curves. These formulas are generally empirically derived and contain the major embrittlement-inducing elements such as copper, nickel and phosphorus. There are a number of such trend curves used in various regulatory guides used in the US, France, Germany, Russia and Japan. These trend curves are often supported by surveillance data and regularly assessed in view of updated surveillance databases. With the recent worldwide move towards life extension of existing reactors above their initially-scheduled lifetime of 40 years, adequate and accurate modeling of irradiation embrittlement becomes a concern for long term operation. The aim of this work is to assess the performance of the embrittlement trend curves used in a regulatory perspective. The work presented here is limited to US and French trend curves because the reactor pressure vessels of the Belgian nuclear power plants are either Westinghouse or Framatome design. The chemical composition of the Belgian RPVs being very close to the one of the French 900 MW units, the French trend curve is used except for the Doel 1-2 units for which these curves are not applicable due to the higher copper content of the welds. In this case, the U.S. trend curves are used. The aim of this work is to evaluate the performance of the embrittlement trend curves used in a regulatory perspective to represent the experimental data obtained in the BR2 reactor. In particular, the French (FIM, FIS) and the US (Reg. Guide 1.99 Rev. 2, ASTM E900-02, EWO and EONY) formulas are of prime interest. The results obtained clearly show that the French trend curves tend to over-estimate the actual irradiation hardening while the US curves under-estimate it. Within the long term operation perspective, both over- and under-estimating are undesirable and therefore the

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

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

    1998-01-01

    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 ∼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

  3. Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

    Watanabe, Katsutoshi

    1987-09-01

    The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

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

    Suter, J. D., E-mail: pradeep.ramuhalli@pnnl.gov; Ramuhalli, P., E-mail: pradeep.ramuhalli@pnnl.gov; Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R. [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); McCloy, J. S., E-mail: john.mccloy@wsu.edu; Xu, K., E-mail: john.mccloy@wsu.edu [Washington State University, PO Box 642920, Pullman, WA 99164 (United States)

    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.

  5. Effects of surface condition on aqueous corrosion and environmental embrittlement of iron aluminides

    Perrin, R.L.; Buchanan, R.A. [Univ. of Tennessee, Knoxville, TN (United States)

    1996-08-01

    Effects of retained high-temperature surface oxides, produced during thermomechanical processing and/or heat treatment, on the aqueous-corrosion and environmental-embrittlement characteristics of Fe{sub 3}Al-based iron aluminides (FA-84, FA-129 and FAL-Mo), a FeAl-based iron aluminide (FA-385), and a disordered low-aluminum Fe-Al alloy (FAPY) were evaluated. All tests were conducted at room temperature in a mild acid-chloride solution. In cyclic-anodic-polarization testing for aqueous-corrosion behavior, the surface conditions examined were: as-received (i.e., with the retained high-temperature oxides), mechanically cleaned and chemically cleaned. For all materials, the polarization tests showed the critical pitting potentials to be significantly lower in the as-received condition than in the mechanically-cleaned and chemically-cleaned conditions. These results indicate detrimental effects of the retained high-temperature oxides in terms of increased susceptibilities to localized corrosion. In 200-hour U-bend stress-corrosion-cracking tests for environmental-embrittlement behavior, conducted at open-circuit corrosion potentials and at a hydrogen-charging potential of {minus}1500 mV (SHE), the above materials (except FA-385) were examined with retained oxides and with mechanically cleaned surfaces. At the open-circuit corrosion potentials, none of the materials in either surface condition underwent cracking. At the hydrogen-charging potential, none of the materials with retained oxides underwent cracking, but FA-84, FA-129 and FAL-Mo in the mechanically cleaned condition did undergo cracking. These results suggest beneficial effects of the retained high-temperature oxides in terms of increased resistance to environmental hydrogen embrittlement.

  6. Hydrogen effect on embrittlement of iron and steel

    Shved, M.M.

    1981-01-01

    Some existing hypothesis brittleness of metals are considered. The following explanation of reversible hydrogen brittleness is suggested: hydrogen presence in iron and steel brings about the increase in the critical shear stress and the yield stress at all stages of plastic deformation (hydrogen, blocking dislocations hinders plastic shears) and the decrease of rupture strength. Decreasing forces of interatomic interaction of the surface layer some scores interatomic distances thick, hydrogen decreases the resistance of normal stresses to its effect. Thus, whatever mechanism brings about the formation of the first cracks in the metal in the presence of absorbed hydrogen, they appear at lower outside applied stresses. In the framework of the model suggested one can explain experimentally observed changes of mechanical properties of iron and steel under hydrogen effect

  7. Lifetime embrittlement of reactor core materials

    Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G.; White, C.J.

    1994-08-01

    Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10 24 n/M 2 (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K IC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K IC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

  8. Effect of Low-Temperature Sensitization on Hydrogen Embrittlement of 301 Stainless Steel

    Chieh Yu; Ren-Kae Shiue; Chun Chen; Leu-Wen Tsay

    2017-01-01

    The effect of metastable austenite on the hydrogen embrittlement (HE) of cold-rolled (30% reduction in thickness) 301 stainless steel (SS) was investigated. Cold-rolled (CR) specimens were hydrogen-charged in an autoclave at 300 or 450 °C under a pressure of 10 MPa for 160 h before tensile tests. Both ordinary and notched tensile tests were performed in air to measure the tensile properties of the non-charged and charged specimens. The results indicated that cold rolling caused the transforma...

  9. Effects of irradiation on mechanical properties

    Server, W.L.; Griesbach, T.J.; Dragunov, Y.; Amaev, A.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The effects of irradiation on the mechanical properties of reactor pressure vessel steels are explained. This chapter provides some background on the critical elements controlling neutron damage effects. Distinction is made between vessels made in the USA and in the former USSR

  10. A study of the mechanisms for the irradiation embrittlement of reactor pressure vessel steels

    Solt, G.; Zimmermann, U.; Waeber, W.B.; Mercier, O.; Frisius, F.; Ghazi-Wakili, K.

    1987-03-01

    Irradiation damage particles were detected by small angle neutron scattering and positron annihilation techniques in two RPV steels. The particle radii were 8A and 14A prior to heat treatments for the plate and weldment, respectively; annealing leads to coarsening in the weldment, the volume fraction remains essentially constant at about 0.14%. The model of copper-rich precipitates 'diluted' by Mn atoms or, alternatively, by vacancy agglomerates is consistent with the neutron scattering data, the presence of simple voids in the weldment would contradict the positron results. Preliminary results on these steels and also on related alloys by methods 'new' in this field are reported. (author)

  11. Effects of strain rate, stress condition and environment on iodine embrittlement of Ziracloy-2

    Une, K.

    1979-01-01

    Iodine stress corrosion cracking (SCC) susceptibility of Zircaloy became higher with decreasing strain rate. Critical strain rate, below which high SCC severity was observed, substantially depended on Zircaloy stress condition. This strain rate (7 x 10 -3 min -1 ) under plane strain condition was about 3.5 times as fast as that (2 x 10 -3 min -1 ) under uniaxial condition. The maximum iodine embrittlement in Zircaloy was found in stress ratio α (axial/tangential stress) range of 0.5 to 0.7. No embrittlement occurred at α = infinity because of its texture effect. The SCC fracture stresses were about 39 kg/mm 2 for unirradiated and stress-relieved material, and about 34 kg/mm 2 for recrystallized material, whose ratios to yield strength of each material were 0.8 and 1.2. Impurity gases of oxygen and moisture in the iodine had the effects of reducing Zircaloy SCC susceptibility. Stress-relieved material was more sensitive to environmental impurities than recrystallized material

  12. Embrittlement data base, version 1

    Wang, J.A.

    1997-08-01

    The aging and degradation of light-water-reactor (LWR) pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel (RPV) materials depends on many different factors such as flux, fluence, fluence spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters and issues such as low-leakage-fuel management, possible life extension, and the need for annealing the pressure vessel. Large amounts of data from surveillance capsules and test reactor experiments, comprising many different materials and different irradiation conditions, are needed to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) is such a comprehensive collection of data resulting from merging version 2 of the Power Reactor Embrittlement Data Base (PR-EDB). Fracture toughness data were also integrated into Version 1 of the EDB. For power reactor data, the current EDB lists the 1,029 Charpy transition-temperature shift data points, which include 321 from plates, 125 from forgoings, 115 from correlation monitor materials, 246 from welds, and 222 from heat-affected-zone (HAZ) materials that were irradiated in 271 capsules from 101 commercial power reactors. For test reactor data, information is available for 1,308 different irradiated sets (352 from plates, 186 from forgoings, 303 from correlation monitor materials, 396 from welds and 71 from HAZs) and 268 different irradiated plus annealed data sets

  13. Radiation embrittlement of metals and alloys

    Wechsler, M.S.

    1975-01-01

    Three types of radiation embrittlement are identified: (1) radiation embrittlement in nominally ductile metals, (2) radiation embrittlement in metals that undergo a ductile-brittle transition, and (3) high-temperature grain boundary embrittlement. This paper deals with type (1) and, more briefly, type (2) radiation embrittlement. Radiation embrittlement in nominally ductile metals is characterized by the premature onset of plastic instability, which causes a sharp decrease in the macroscopic plastic strain that the material can sustain before necking (uniform strain) and breaking (fracture strain). Dislocation channeling seems to be largely responsible and experimental results are reviewed. The origin of dislocation channeling is discussed. Irradiated metals that exhibit a ductile-brittle transition show an increase in the transition temperature but the nature of the transition (shear to cleavage fracture) does not appear to be greatly altered. A key factor is the temperature dependence of yielding and how it is affected upon irradiation. Impurities exert an influence on the stability of radiation-produced defect clusters and thus can alter the amount of radiation embrittlement experienced upon irradiation at somewhat elevated temperatures. In general, radiation embrittlement appears to stem mostly from changes in plastic properties (particularly in the trend toward more dynamic and inhomogeneous plastic deformation) rather than from changes in the inherent fracture process. 63 references, 10 figures

  14. Low temperature radiation embrittlement for reactor vessel steels

    Ginding, I.A.; Chirkina, L.A.

    1978-01-01

    General conceptions of cold brittleness of bcc metals are in a review. Considered are experimental data and theoretical representations about the effect of irradiation conditions, chemical composition, phase and structural constitutions, grain size, mechanical and thermomechanical treatments on low-temperature irradiation embrittlement of reactor vessel steels. Presented are the methods for increasing radiation stability of metals (carbon and Cr-Mo steels) used in manufacturing reactor vessels

  15. Overview of French activities on neutron radiation embrittlement of pressure vessel steel

    Brillaud, C [Electricite de France (EDF), 37 - Tours (France); Keroulas, F de [Electricite de France (EDF), 93 - Saint-Denis (France); Pichon, C [Electricite de France (EDF), 69 - Villeurbanne (France); Teissier, A [Electricite de France (EDF), 92 - Courbevoie (France). Service Etudes et Projets Thermiques et Nucleaires

    1994-12-31

    This paper describes recent developments in France`s pressure vessel surveillance program, particularly aimed at assessing the irradiation-caused embrittlement of EDF`s PWRs. The first part presents surveillance program results for base metal, weld metal and heat-affected zones for 74 capsules removed from 34 units. Fluence ranges from 0.3.10{sup 19} n.cm{sup -2} to 5.5.10{sup 19} n.cm{sup -2}. The second part considers research and development activities in this area: these include the metallurgical structure effects of segregated bands on mechanical properties and the embrittlement rate under irradiation, as well as the effect of irradiation parameters such as flux and neutron spectrum on irradiation embrittlement, and more especially to obtain the best damage assessment. (authors). 14 refs., 5 figs., 1 tab.

  16. Investigations of low-temperature neutron embrittlement of ferritic steels

    Farrell, K.; Mahmood, S.T.; Stoller, R.E.; Mansur, L.K.

    1992-01-01

    Investigations were made into reasons for accelerated embrittlement of surveillance specimens of ferritic steels irradiated at 50C at the High Flux Isotope Reactor (HFIR) pressure vessel. Major suspects for the precocious embrittlement were a highly thermalized neutron spectrum,a low displacement rate, and the impurities boron and copper. None of these were found guilty. A dosimetry measurement shows that the spectrum at a major surveillance site is not thermalized. A new model of matrix hardening due to point defect clusters indicates little effect of displacement rate at low irradiation temperature. Boron levels are measured at 1 wt ppM or less, inadequate for embrittlement. Copper at 0.3 wt % and nickel at 0.7 wt % are shown to promote radiation strengthening in iron binary alloys irradiated at 50 to 60C, but no dependence on copper and nickel was found in steels with 0.05 to 0.22% Cu and 0.07 to 3.3% Ni. It is argued that copper impurity is not responsible for the accelerated embrittlement of the HFIR surveillance specimens. The dosimetry experiment has revealed the possibility that the fast fluence for the surveillance specimens may be underestimated because the stainless steel monitors in the surveillance packages do not record an unexpected component of neutrons in the spectrum at energies just below their measurement thresholds of 2 to 3 MeV

  17. The Test Reactor Embrittlement Data Base (TR-EDB)

    Stallmann, F.W.; Kam, F.B.K.; Wang, J.A.

    1993-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is part of an ongoing program to collect test data from materials irradiations to aid in the research and evaluation of embrittlement prediction models that are used to assure the safety of pressure vessels in power reactors. This program is being funded by the US Nuclear Regulatory Commission (NRC) and has resulted in the publication of the Power Reactor Embrittlement Data Base (PR-EDB) whose second version is currently being released. The TR-EDB is a compatible collection of data from experiments in materials test reactors. These data contain information that is not obtainable from surveillance results, especially, about the effects of annealing after irradiation. Other information that is only available from test reactors is the influence of fluence rates and irradiation temperatures on radiation embrittlement. The first version of the TR-EDB will be released in fall of 1993 and contains published results from laboratories in many countries. Data collection will continue and further updates will be published

  18. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  19. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  20. Status and task of the study on the hydrogen embrittlement of zirconium alloys

    Nagase, Fumihisa; Furuta, Teruo; Seino, Shun; Komatsu, Kazushi.

    1995-08-01

    As the burnup of the LWR fuel is extended, waterside corrosion and hydrogen pickup increase in the Zircaloy cladding. Hydrogen embrittlement of Zircaloy is one of the main factors which may limit the life of the fuel rod. This report presents a review on the hydrogen embrittlement of zirconium and its alloys including the irradiated materials. Research tasks for the reduction of ductility in the high burnup fuel cladding are also discussed. Many fundamental investigations have been performed on the hydrogen embrittlement of zirconium alloys. However, the embrittlement mechanism of the high burnup fuel cladding is complicated. Especially, a coupled effect of hydrides and radiation defects are expected to be pronounced with neutron dose increase. In order to evaluate the reduction of ductility of the higher burnup fuel cladding properly, it is necessary to investigate the coupled effect of these two factors by systematic examinations. (author) 64 refs

  1. Nuclear power plant life extension and management aspects; neutron irradiation embrittlement and stress corrosion cracking - two possible degradation mechanisms and methods for their mitigation

    Tipping, P.; Ineichen, U.; Cripps, R.C.

    1994-01-01

    The response of a mock-up low alloy ferritic reactor pressure vessel (RPV) steel and associated weldments to neutron irradiation has been studied using a combination of hardness, tensile, fracture mechanical and toughness tests in combination with annealing treatments. Thermal analysis using isochronal and isothermal techniques has indicated that annealing at a minimum of 440 o C for 168h is needed to mitigate neutron embrittlement received at 290 o C. Rates of re-embrittlement after annealing and reirradiating are no faster than initial rates, even up to neutron fluences as high as 5x10 19 cm -2 (energy E>1 MeV). All mechanical properties measured benefited from annealing. Thus, annealing is indicated as one measure for maintaining mechanical properties in irradiated low alloy steels and welds and should be considered in plant life management strategies. The influence of simulated reactor coolant water chemistry on the stress corrosion cracking propensity of ferritic low alloy steel specimens in autoclave loop experiments has also been studied. The double cantilever bend specimens were fatigue pre-cracked and wedge-loaded to different degrees to induce nominal stress intensity factors between 15-95 MPa.m 1/2 . Other specimens were subjected to stress using a tensile loading device integral with the test autoclave. The importance of close control of the dissolved oxygen content and the conductivity of the water has become evident under these experimental conditions. The RPV material and degree and mode of loading are also important parameters in SCC studies; stress intensity factors above 30 MPa.m 1/2 have been associated with SCC in these studies. (author) 2 figs., 13 refs

  2. Effects of metallurgical variables on hydrgen embrittlement in types 316, 321, and 347 stainless steels

    Rozenak, P.; Eliezer, D.

    1984-01-01

    Hydrogen embrittlement of 316, 321 and 347 types austenitic stainless steels has been studied by charging thin tensile specimens with hydrogen through cathodic polarization. Throughout this study we have compared solution annealed samples having various prior austenitic grain-size with samples given the additional sensitization treatment. The results show that refined grains improves the resistance to hydrogen cracking regardless of the failure mode. The sensitized specimens were predominantly intergranular, while the annealed specimens show massive regions of microvoid coalescence producing ductile rupture. 347 type stainless steel is much more susceptible to hydrogen embrittlement than 321 type steel, and 316 type is the most resistant to hydrogen embrittlement. the practical implication of the experimental conclusions are discussed

  3. Effect of aluminium concentration and boron dopant on environmental embrittlement in FeAl aluminides

    Liu, C.T.; George, E.P.

    1991-01-01

    This paper reports on the room-temperature tensile properties of FeAl aluminides determined as functions of aluminum concentration (35 to 43 at. % Al), test environment, and surface (oil) coating. The two lower aluminum alloys containing 35 and 36.5% Al are prone to severe environmental embrittlement, while the two higher aluminum alloys with 40 and 43% Al are much less sensitive to change in test environment and surface coating. The reason for the different behavior is that the grain boundaries are intrinsically weak in the higher aluminum alloys, and these weak boundaries dominate the low ductility and brittle fracture behavior of the 40 and 43% Al alloys. When boron is added to the 40% Al alloy as a grain-boundary strengthener, the environmental effect becomes prominent. In this case, the tensile ductility of the boron-doped alloy, just like that of the lower aluminum alloys, can be dramatically improved by control of test environment (e.g. dry oxygen vs air). Strong segregation of boron to the grain boundaries, with a segregation factor of 43, was revealed by Auger analyses

  4. Effect of Low-Temperature Sensitization on Hydrogen Embrittlement of 301 Stainless Steel

    Chieh Yu

    2017-02-01

    Full Text Available The effect of metastable austenite on the hydrogen embrittlement (HE of cold-rolled (30% reduction in thickness 301 stainless steel (SS was investigated. Cold-rolled (CR specimens were hydrogen-charged in an autoclave at 300 or 450 °C under a pressure of 10 MPa for 160 h before tensile tests. Both ordinary and notched tensile tests were performed in air to measure the tensile properties of the non-charged and charged specimens. The results indicated that cold rolling caused the transformation of austenite into α′ and ε-martensite in the 301 SS. Aging at 450 °C enhanced the precipitation of M23C6 carbides, G, and σ phases in the cold-rolled specimen. In addition, the formation of α′ martensite and M23C6 carbides along the grain boundaries increased the HE susceptibility and low-temperature sensitization of the 450 °C-aged 301 SS. In contrast, the grain boundary α′-martensite and M23C6 carbides were not observed in the as-rolled and 300 °C-aged specimens.

  5. The effects of composition on the environmental embrittlement of Fe{sub 3}Al alloys

    Alven, D.A.; Stoloff, N.S. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    1997-12-01

    This paper reviews recent research on embrittlement of iron aluminides at room temperature brought about by exposure to moisture or hydrogen. The tensile and fatigue crack growth behavior of several Fe-28Al-5Cr alloys with small additions of Zr and C are described. It will be shown that fatigue crack growth behavior is dependent on composition, environment, humidity level, and frequency. Environments studied include vacuum, oxygen, hydrogen gas, and moist air. All cases of embrittlement are ultimately traceable to the interaction of hydrogen with the crack tip.

  6. Effect of microstructure on the susceptibility of a 533 steel to temper embrittlement

    Raoul, S.; Marini, B.; Pineau, A.

    1998-01-01

    In ferritic steels, brittle fracture usually occurs at low temperature by cleavage. However the segregation of impurities (P, As, Sn etc..) along prior γ grain boundaries can change the brittle fracture mode from transgranular to intergranular. In quenched and tempered steels, this segregation is associated with what is called the temper-embrittlement phenomenon. The main objective of the present study is to investigate the influence of the as-quenched microstructure (lower bainite or martensite) on the susceptibility of a low alloy steel (A533 cl.1) to temper-embrittlement. Dilatometric tests were performed to determine the continous-cooling-transformation (CCT) diagram of the material and to measure the critical cooling rate (V c ) for a martensitic quench. Then subsized Charpy V-notched specimens were given various cooling rates from the austenitization temperature to obtain a wide range of as-quenched microstructures, including martensite and bainite. These specimens were subsequently given a heat treatment to develop temper embrittlement and tested to measure the V-notch fracture toughness at -50 C. The fracture surfaces were examined by SEM. It is shown that martensitic microstructures are more susceptible to intergranular embrittlement than bainitic microstructures. These observed microstructural influences are briefly discussed. (orig.)

  7. Effect of microstructure on the susceptibility of a 533 steel to temper embrittlement

    Raoul, S.; Marini, B. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Pineau, A. [CNRS, Evry (France). Centre de Materiaux

    1998-11-01

    In ferritic steels, brittle fracture usually occurs at low temperature by cleavage. However the segregation of impurities (P, As, Sn etc..) along prior {gamma} grain boundaries can change the brittle fracture mode from transgranular to intergranular. In quenched and tempered steels, this segregation is associated with what is called the temper-embrittlement phenomenon. The main objective of the present study is to investigate the influence of the as-quenched microstructure (lower bainite or martensite) on the susceptibility of a low alloy steel (A533 cl.1) to temper-embrittlement. Dilatometric tests were performed to determine the continous-cooling-transformation (CCT) diagram of the material and to measure the critical cooling rate (V{sub c}) for a martensitic quench. Then subsized Charpy V-notched specimens were given various cooling rates from the austenitization temperature to obtain a wide range of as-quenched microstructures, including martensite and bainite. These specimens were subsequently given a heat treatment to develop temper embrittlement and tested to measure the V-notch fracture toughness at -50 C. The fracture surfaces were examined by SEM. It is shown that martensitic microstructures are more susceptible to intergranular embrittlement than bainitic microstructures. These observed microstructural influences are briefly discussed. (orig.) 11 refs.

  8. Effect of solute interaction on interfacial and grain boundary embrittlement in binary alloys

    Lejček, Pavel

    2013-01-01

    Roč. 48, č. 6 (2013), 2574-2580 ISSN 0022-2461 R&D Projects: GA ČR GAP108/12/0144 Institutional research plan: CEZ:AV0Z10100520 Keywords : interfacial segregation * grain boundary embrittlement * binary interaction * modeling * thermodynamics Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.305, year: 2013

  9. Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

    1998-10-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

  10. Embrittlement of the Shippingport reactor shield tank

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

    1989-01-01

    Surveillance specimens from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory showed an unexpectedly high degree of embrittlement relative to the data obtained on similar materials in Materials Testing Reactors (MTRs). The results suggest a possible negative flux effect and raise the issue of embrittlement of the pressure vessel support structures of commercial light water reactors. To help resolve this issues, a program was initiated to characterize the irradiation embrittlement of the neutron shield tank (NST) from the decommissioned Shippingport reactor. The Shippingport NST operated at 55 degree C (130 degree F) and was fabricated from A212 Grade B steel, similar to the vessel material in HFIR. The inner wall of the NST was exposed to a total maximum fluence of ∼ 6 x 10 17 n/cm 2 (E > 1 MeV) over a life of 9.25 effective full power years. This corresponds to a fast flux of 2.1 x 10 9 n/cm 2 x s and is comparable to the conditions for the HFIR surveillance specimens. The results indicate that irradiation increases the 15 ft x lb Charpy transition temperature (CTT) by ∼25 degree C (45 degree F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens and is consistent with that expected from the MTR data base. However, the actual value of CTT is high, and the toughness at service temperature is low, even when compared with the HFIR data. The increase in yield stress is ∼50 MPa, which is comparable to the HFIR data. The results also indicate a lower impact strength and higher transition temperature for the TL orientation than that for the LT orientation. Some effects of the location across the thickness of the wall are also observed for the LT specimens; CTT is slightly greater for the specimens from the inner region of the wall

  11. A cohesive zone model to simulate the hydrogen embrittlement effect on a high-strength steel

    G. Gobbi

    2016-01-01

    Full Text Available The present work aims to model the fracture mechanical behavior of a high-strength low carbon steel, AISI 4130 operating in hydrogen contaminated environment. The study deals with the development of 2D finite element cohesive zone model (CZM reproducing a toughness test. Along the symmetry plane over the crack path of a C(T specimen a zero thickness layer of cohesive elements are implemented in order to simulate the crack propagation. The main feature of this kind of model is the definition of a traction-separation law (TSL that reproduces the constitutive response of the material inside to the cohesive elements. Starting from a TSL calibrated on hydrogen non-contaminated material, the embrittlement effect is simulated by reducing the cohesive energy according to the total hydrogen content including the lattice sites (NILS and the trapped amount. In this perspective, the proposed model consists of three steps of simulations. First step evaluates the hydrostatic pressure. It drives the initial hydrogen concentration assigned in the second step, a mass diffusion analysis, defining in this way the contribution of hydrogen moving across the interstitial lattice sites. The final stress analysis, allows getting the total hydrogen content, including the trapped amount, and evaluating the new crack initiation and propagation due to the hydrogen presence. The model is implemented in both plane strain and plane stress configurations; results are compared in the discussion. From the analyses, it resulted that hydrogen is located only into lattice sites and not in traps, and that the considered steel experiences a high hydrogen susceptibility. By the proposed procedure, the developed numerical model seems a reliable and quick tool able to estimate the mechanical behavior of steels in presence of hydrogen.

  12. Models for embrittlement recovery due to annealing of reactor pressure vessel steels

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1995-05-01

    The reactor pressure vessel (RPV) surrounding the core of a commercial nuclear power plant is subject to embrittlement due to exposure to high energy neutrons. The effects of irradiation embrittlement can be reduced by thermal annealing at temperatures higher than the normal operating conditions. However, a means of quantitatively assessing the effectiveness of annealing for embrittlement recovery is needed. The objective of this work was to analyze the pertinent data on this issue and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. Data were gathered from the Test Reactor Embrittlement Data Base and from various annealing reports. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Independent variables considered in the analysis included material chemistries, annealing time and temperature, irradiation time and temperature, fluence, and flux. To identify important variables and functional forms for predicting embrittlement recovery, advanced statistical techniques, including pattern recognition and transformation analysis, were applied together with current understanding of the mechanisms governing embrittlement and recovery. Models were calibrated using multivariable surface-fitting techniques. Several iterations of model calibration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness data produced correlation models for estimating Charpy upper shelf energy and transition temperature after irradiation and annealing. This work provides a clear demonstration that (1) microhardness recovery is generally a very good surrogate for shift recovery, and (2) there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

  13. Solubility of hydrogen in metals and its effect of pore-formation and embrittlement. Ph.D. Thesis

    Shahani, H. R.

    1984-01-01

    The effect of alloying elements on hydrogen solubility were determined by evaluating solubility equations and interaction coefficients. The solubility of dry hydrogen at one atmosphere was investigated in liquid aluminum, Al-Ti, Al-Si, Al-Fe, liquid gold, Au-Cu, and Au-Pd. The design of rapid heating and high pressure casting furnaces used in meta foam experiments is discussed as well as the mechanism of precipitation of pores in melts, and the effect of hydrogen on the shrinkage porosity of Al-Cu and Al-Si alloys. Hydrogen embrittlement in iron base alloys is also examined.

  14. Effect of ternary solute interaction on interfacial segregation and grain boundary embrittlement

    Lejček, Pavel

    2013-01-01

    Roč. 48, č. 14 (2013), 4965-4972 ISSN 0022-2461 R&D Projects: GA MŠk(CZ) LM2011026; GA ČR GAP108/12/0144 Institutional research plan: CEZ:AV0Z10100520 Keywords : interfacial segregation * intergranular embrittlement * solute interaction * modeling * thermodynamics Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 2.305, year: 2013

  15. Effect of neutron irradiation on vanadium alloys

    Braski, D.N.

    1986-01-01

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

  16. Effect of neutron irradiation on vanadium alloys

    Braski, D.N.

    1986-01-01

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

  17. Analysis of the surveillance test data on irradiation embrittlement of the reactor pressure vessel steels in LWRs

    Lee, Gyoeng Geun; Kwon, Jun Hyun

    2010-11-01

    The surveillance test data in Korean LWRs were analyzed from a viewpoint of materials science. TTS change with the neutron irradiation were compared to the model values of the RG1.99/2 and NUREG/CR-6551. The model values of TTS were higher than the actual values of TTS. It was impossible to find a relationship between TTS and neutron fluence in weld data. The correlation of the increase in YS (yield strength) and TTS with neutron irradiation was also investigated. Like the result of TTS change, the YS/TTS showed the correlations in plate/forgings metals, however no correlation in weld metals. The data were similar to Odette's result about US surveillance tests. From the empirical relationships, the TTS curve change could be predicted using the CVN test result of the unirradiated specimen and the change in YS with neutron irradiation of the specimen

  18. Cladding embrittlement during postulated loss-of-coolant accidents.

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

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

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    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.

  20. The study of the irradiation-induced embrittlement of reactor pressure vessels. Analysis of surveillance test specimens of a commercial nuclear reactor pressure vessel studied by three-dimensional atom probe and positron annihilation

    Nagai, Yasuyoshi; Toyama, Takeshi; Hasegawa, Masayuki

    2007-01-01

    The study of embrittlement of nuclear power reactor pressure vessels (RPVs) is of critical importance for the safety assessment in the nuclear industry. Some origins of embrittlement are attributed to fine Cu precipitates, matrix defects, grain boundary segregation of P and late blooming phase. This review article described nanostructural observation by three-dimensional atom probe (3DAP) and positron annihilation spectroscopy (PAS). The density and sizes of Cu-rich nanoprecipitates and grain boundary segregation are sensitively detected by 3DAP, and vacancies are probed by PAS. Element analysis around vacancies and fine microstructural Cu precipitates not containing vacancies are successfully observed by a coincidence doppler broadening method. The nanostructural evolution of irradiation-induced Cu-rich nanoprecipitates (CRNPs) and vacancy clusters in surveillance test specimens of commercial nuclear reactor pressure vessel steel welds of Doel-2 in Belgium were revealed by combining 3DAP and PAS. In both medium (0.13 wt%) and high (0.30 wt%) Cu welds, the CRNPs were found to form readily at the very beginning of the reactor lifetime. On the other hand, small vacancy clusters start appearing after the initial Cu precipitates and accumulate steadily with increasing neutron dose. The CRNPs were also observed at very low dose rate of neutrons in the test specimen of Calder Hall Reactor of Japan Atomic Power Company. The significant enhancement of these Cu precipitates results in the embrittlement in practical RPVs. At very high dose of 2.2x10 18 n/cm 2 by JMTR, the Cu precipitates were scarcely observed, and the irradiation-induced embrittlement was primarily caused from vacancy-impurity complexes and dislocation loops. (author)

  1. High temperature embrittlement of metals by helium

    Schroeder, H.

    1983-01-01

    The present knowledge of the influence of helium on the high temperature mechanical properties of metals to be used as structural materials in fast fission and in future fusion reactors is reviewed. A wealth of experimental data has been obtained by many different experimental techniques, on many different alloys, and on different properties. This review is mostly concentrated on the behaviour of austenitic alloys -especially austenitic stainless steels, for which the data base is by far the largest - and gives only a few examples of special bcc alloys. The effect of the helium embrittlement on the different properties - tensile, fatigue and, with special emphasis, creep - is demonstrated by representative results. A comparison between data obtained from in-pile (-beam) experiments and from post-irradiation (-implantation) experiments, respectively, is presented. Theoretical models to describe the observed phenomena are briefly outlined and some suggestions are made for future work to resolve uncertainties and differences between our experimental knowledge and theoretical understanding of high temperature helium embrittlement. (author)

  2. Power Reactor Embrittlement Data Base

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1990-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: (1) to compile and to verify the quality of the PR-EDB; (2) to provide user-friendly software to access and process the data; (3) to explore or confirm embrittlement prediction models; and (4) to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. To achieve these goals, the data base architecture was designed after much discussion and planning with prospective users, namely, material scientists and members of the research staff. The current compilation of the PR-EDB (Version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points for 110 different irradiated base materials and 161 data points for 79 different welds. Results from heat-affected zone materials are also listed. The time and effort required to process and evaluate different types of data in the PR-EDB have been drastically reduced from previous data bases. The Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of PR-EDB and will be supplementing the data base with additional data and documentation

  3. Effect of pre-strain on susceptibility of Indian Reduced Activation Ferritic Martensitic Steel to hydrogen embrittlement

    Sonak, Sagar; Tiwari, Abhishek; Jain, Uttam; Keskar, Nachiket; Kumar, Sanjay; Singh, Ram N.; Dey, Gautam K.

    2015-01-01

    The role of pre-strain on hydrogen embrittlement susceptibility of Indian Reduced Activation Ferritic Martensitic Steel was investigated using constant nominal strain-rate tension test. The samples were pre-strained to different levels of plastic strain and their mechanical behavior and mode of fracture under the influence of hydrogen was studied. The effect of plastic pre-strain in the range of 0.5–2% on the ductility of the samples was prominent. Compared to samples without any pre-straining, effect of hydrogen was more pronounced on pre-strained samples. Prior deformation reduced the material ductility under the influence of hydrogen. Up to 35% reduction in the total strain was observed under the influence of hydrogen in pre-strained samples. Hydrogen charging resulted in increased occurrence of brittle zones on the fracture surface. Hydrogen Enhanced Decohesion (HEDE) was found to be the dominant mechanism of fracture.

  4. A new method for improving the reliability of fracture toughness surveillance of nuclear pressure vessel by neutron irradiated embrittlement

    Zhang Xinping; Shi Yaowu

    1992-01-01

    In order to obtain more information from neutron irradiated sample specimens and raise the reliability of fracture toughness surveillance test, it has more important significance to repeatedly exploit the broken Charpy-size specimen which had been tested in surveillance test. In this work, on the renewing design and utilization for Charpy-size specimens, 9 data of fracture toughness can be gained from one pre-cracked side-grooved Charpy-size specimen while at the preset usually only 1 to 3 data of fracture toughness can be obtained from one Chharpy-size specimen. Thus, it is found that the new method would obviously improve the reliability of fracture toughness surveillance test and evaluation. Some factors which affect the reasonable design of pre-cracked deep side-groove Charpy-size compound specimen have been discussed

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

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

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

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

    1997-01-01

    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 ∼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 shiftclose 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 ∼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

  7. Hydrogen environment embrittlement

    Donovan, J.A.

    1975-01-01

    Exposure of many metals to gaseous hydrogen causes losses in elongation, reduction of area, and fracture toughness, and causes increases in slow crack growth rate or fatigue life compared with values obtained in air or vacuum. Hydrogen pressure, temperature, and purity significantly influence deleterious effects. The strength and structural characteristics of the metal influence the degradation of its properties by hydrogen. Several theories have been proposed to explain the loss of properties in hydrogen, but none has gained wide acceptance. The embrittlement mechanism and the role of diffusion are, therefore, open questions and need more quantitative experimental data both to test the proposed theories and to allow the development of realistic preventive measures. (U.S.)

  8. Effect of post weld heat treatments on the resistance to the hydrogen embrittlement of soft martensitic stainless steel

    Hazarabedian, Alfredo; Ovejero Garcia, Jose; Bilmes, P.; Llorente, C.

    2003-01-01

    The effect of external hydrogen on the tensile properties of an all weld sample of a soft martensitic stainless steel was studied. The material was tested in the as weld condition and after tempered conditions modifying the austenite content, and changing the quantity, type and distribution of precipitates. Hydrogen was introduced by cathodic charge or by immersion in an acid brine saturated whit 1 atm hydrogen sulphide, during the mechanical test. The as weld condition showed a good resistance in the hydrogen sulphide, were the tempered samples were embrittled. Under cathodic charge, all samples were susceptible to hydrogen damage. The embritting mechanisms were the same in both environments. When the austenite content, was below 10% the crack path is on the primary austenite grain boundary. At higher austenite content, the crack is transgranular. (author)

  9. Pressure vessel steels: influence of chemical composition on irradiation sensitivity

    Ghoniem, M.M.; Hammad, F.H.

    1998-01-01

    Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

  10. Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

    Krasikov, E.A.; Nikolaenko, V.A.

    2008-01-01

    Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

  11. Evaluation of irradiation damage effect by applying electric properties based techniques

    Acosta, B.; Sevini, F.

    2004-01-01

    The most important effect of the degradation by radiation is the decrease in the ductility of the pressure vessel of the reactor (RPV) ferritic steels. The main way to determine the mechanical behaviour of the RPV steels is tensile and impact tests, from which the ductile to brittle transition temperature (DBTT) and its increase due to neutron irradiation can be calculated. These tests are destructive and regularly applied to surveillance specimens to assess the integrity of RPV. The possibility of applying validated non-destructive ageing monitoring techniques would however facilitate the surveillance of the materials that form the reactor vessel. The JRC-IE has developed two devices, focused on the measurement of the electrical properties to assess non-destructively the embrittlement state of materials. The first technique, called Seebeck and Thomson Effects on Aged Material (STEAM), is based on the measurement of the Seebeck coefficient, characteristic of the material and related to the microstructural changes induced by irradiation embrittlement. With the same aim the second technique, named Resistivity Effects on Aged Material (REAM), measures instead the resistivity of the material. The purpose of this research is to correlate the results of the impact tests, STEAM and REAM measurements with the change in the mechanical properties due to neutron irradiation. These results will make possible the improvement of such techniques based on the measurement of material electrical properties for their application to the irradiation embrittlement assessment

  12. Comparison of embrittlement trend curves to high fluence surveillance results

    Bogaert, A.S.; Gerard, R.; Chaouadi, R.

    2011-01-01

    In the regulatory justification of the integrity of the reactor pressure vessels (RPV) for long term operation, use is made of predictive formulas (also called trend curves) to evaluate the RPV embrittlement (expressed in terms of RTNDT shifts) in function of fluence, chemical composition and in some cases temperature, neutron flux or product form. It has been shown recently that some of the existing or proposed trend curves tend to underpredict high dose embrittlement. Due to the scarcity of representative surveillance data at high dose, some test reactor results were used in these evaluations and raise the issue of representativeness of the accelerated test reactor irradiations (dose rate effects). In Belgium the surveillance capsules withdrawal schedule was modified in the nineties in order to obtain results corresponding to 60 years of operation or more with the initial surveillance program. Some of these results are already available and offer a good opportunity to test the validity of the predictive formulas at high dose. In addition, advanced surveillance methods are used in Belgium like the Master Curve, increased tensile tests, and microstructural investigations. These techniques made it possible to show the conservatism of the regulatory approach and to demonstrate increased margins, especially for the first generation units. In this paper the surveillance results are compared to different predictive formulas, as well as to an engineering hardening model developed at SCK.CEN. Generally accepted property-to-property correlations are critically revisited. Conclusions are made on the reliability and applicability of the embrittlement trend curves. (authors)

  13. Helium embrittlement model and program plan for weldability of ITER materials

    Louthan, M.R. Jr.; Kanne, W.R. Jr.; Tosten, M.H.; Rankin, D.T.; Cross, B.J.

    1997-02-01

    This report presents a refined model of how helium embrittles irradiated stainless steel during welding. The model was developed based on experimental observations drawn from experience at the Savannah River Site and from an extensive literature search. The model shows how helium content, stress, and temperature interact to produce embrittlement. The model takes into account defect structure, time, and gradients in stress, temperature and composition. The report also proposes an experimental program based on the refined helium embrittlement model. A parametric study of the effect of initial defect density on the resulting helium bubble distribution and weldability of tritium aged material is proposed to demonstrate the roll that defects play in embrittlement. This study should include samples charged using vastly different aging times to obtain equivalent helium contents. Additionally, studies to establish the minimal sample thickness and size are needed for extrapolation to real structural materials. The results of these studies should provide a technical basis for the use of tritium aged materials to predict the weldability of irradiated structures. Use of tritium charged and aged material would provide a cost effective approach to developing weld repair techniques for ITER components

  14. Status of reactor pressure vessel embrittlement study in Japan

    Sasajima, H.

    1997-01-01

    Since the construction of Japanese first commercial nuclear power plant in 1966, 52 nuclear power plants have been commissioned in Japan to commercial operation. Japanese first nuclear power plant has now been service for 30 years and the aging of nuclear power plants is steadily progressing in general. Under these circumstances, the Japan Power Engineering and Inspection Corporation (JAPEIC) is executing, under consignment by the Ministry of International Trade and Industry (MITI), the development and verification test programs for plant integrity evaluation technology by which nuclear power plant aging can be appropriately handled. This paper shows the outline of study dealing with embrittlement of RPV caused by neutron irradiation, as one of the activity of JAPEIC. The embrittlement of RPV caused by neutron irradiation is manifested as a shift of transition temperature and as a reduction in Upper Shelf Energy (USE). In JAPEIC, the study dealing with a shift of transition temperature was conducted in the ''Reactor Pressure Vessel Pressurized Thermal Shock Test Project (the PTS Project)'', and the study dealing with a reduction in USE has been conducted in the ''Nuclear Power Plant Life Management Technology (the PLIM Project)''. And the reconstitution technology of surveillance test specimen has been conducted in PLIM Project as one of the measures to improve monitoring above material characteristic changes. The integrity evaluation under the Pressurized Thermal Shock (PTS) events including the effect of neutron irradiation embrittlement was initiated in 1983 FY as the PTS Project and was completed in the 1991 FY. The study verified that plant integrity could be assured at not only the end of design life, but also an extended service life even when the severest PTS events were postulated. The PLIM Project, designed to develop and verify the integrity evaluation technology dealing with reduction of USE by neutron irradiation, was started in the 1996 FY as a 10

  15. Irradiation effects on aluminium and beryllium

    Bieth, M.

    1992-01-01

    ductility of 1.6%. Besides, due to the effects of embrittlement and swelling induced by irradiation, the HFR beryllium reflector elements had to be replaced after more than 25 years of operation. Operational and practical experiences with these reflector elements are commented, as well as main engineering features of the new reflector elements: upper-end fittings of both filler element and insert in stainless steel, no radially drilled holes and no roll pins

  16. An internal-friction study of reactor-pressure-vessel steel embrittlement

    Ouytsel, K. van; Fabry, A.; Batist, R. de; Schaller, R.

    1997-01-01

    Within an enhanced commercial surveillance strategy, the nuclear-research institute SCK.CEN in Mol, Belgium is investigating, by means of internal friction, the microstructural processes responsible for embrittlement of pressure-vessel steels. The experiments were carried out using a torsion pendulum at the Ecole Polytechnique Federale de Lausanne in Switzerland. Amplitude-independent internal-friction experiments teach us that neutron irradiation induces defects which interact with mobile dislocations. Thermal ageing of JRQ and Doel-IV steel does not cause major embrittlement effects. Amplitude-dependent internal-friction experiments allow us to determine a critical amplitude which corresponds to the yield stress of the material as obtained from static tensile tests. The results also correspond to a three-component model for the yield strength taking into account both hardening and non-hardening embrittlement. Investigations of Doel-I-II weld material in different conditions reveal that embrittlement due to irradiation or thermal ageing can be interpreted in terms of a fine interplay between long- and short-range phenomena. (author)

  17. The effects of normal paraffins mobilizers on irradiated polypropylene

    Chen Wenxiu; Gao Ling

    1995-01-01

    The n-paraffins blended with polypropylene (PP) as mobilizer had been investigated. The effectiveness of mobilizer (n-paraffins) on irradiated polypropylene is dependent on the molecular weight of mobilizer and its content on polypropylene. The n-docosame (n-C 22 ) possesses the best effectiveness of radiation tolerance on PP among the mobilizer paraffins: n-decane (n-C 10 ), n-hexadecane (n-C 16 ), n-docosane (n-C 22 ) and n-hexatriacontane (n-C 36 ). The 2% (w/w) content of a given mobilizer is the most effective at reducing the embrittlement of irradiated PP as evidenced by the elongation at break. The physical properties of polypropylene with mobilizers such as density, Young's modulus, the Fraction of free volume and the weight swelling ratio in p-xylene at room temperature were measured. Above phenomena are related with the constructive of blended PP and demonstrated by its physical properties

  18. Effect of microstructure on the impact toughness and temper embrittlement of SA508Gr.4N steel for advanced pressure vessel materials.

    Yang, Zhiqiang; Liu, Zhengdong; He, Xikou; Qiao, Shibin; Xie, Changsheng

    2018-01-09

    The effect of microstructure on the impact toughness and the temper embrittlement of a SA508Gr.4N steel was investigated. Martensitic and bainitic structures formed in this material were examined via scanning electron microscopy, electron backscatter diffraction, transmission electron microscopy, and Auger electron spectroscopy (AES) analysis. The martensitic structure had a positive effect on both the strength and toughness. Compared with the bainitic structure, this structure consisted of smaller blocks and more high-angle grain boundaries (HAGBs). Changes in the ultimate tensile strength and toughness of the martensitic structure were attributed to an increase in the crack propagation path. This increase resulted from an increased number of HAGBs and refinement of the sub-structure (block). The AES results revealed that sulfur segregation is higher in the martensitic structure than in the bainitic structure. Therefore, the martensitic structure is more susceptible to temper embrittlement than the bainitic structure.

  19. Effect of retained austenite stability and morphology on the hydrogen embrittlement susceptibility in quenching and partitioning treated steels

    Zhu, Xu [State Key Lab of Metal Matrix Composites, School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Collaborative Innovation Center for Advanced Ship and Deep-Sea Exploration, Shanghai Jiao Tong University, Shanghai 200240 (China); Zhang, Ke [School of Materials Science and Engineering, University of Shanghai for Science and Technology, Shanghai 200093 (China); Li, Wei, E-mail: weilee@sjtu.edu.cn [State Key Lab of Metal Matrix Composites, School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Jin, Xuejun, E-mail: jin@sjtu.edu.cn [Collaborative Innovation Center for Advanced Ship and Deep-Sea Exploration, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2016-03-21

    The effect of retained austenite (RA) stability and morphology on the hydrogen embrittlement (HE) susceptibility were investigated in a high strength steel subjected to three different heat treatments, i.e., the intercritical annealing quenching and partitioning (IAQP), quenching and partitioning (QP) and quenching and tempering (QT). IAQP treatment results in the coexistence of blocky and filmy morphologies and both QP and QT treatments lead to only filmy RA. No martensitic transformation occurs in QT steel during deformation, while the QP and IAQP undergo the transformation with the same extent. It is shown that the HE susceptibility increases in the following order: QT, QP and IAQP. Despite of the highest strength level and the highest hydrogen diffusion rate, the QT steel is relative immune to HE, suggesting that the metastable RA which transforms to martensite during deformation is detrimental to the HE resistance. The improved resistance to HE by QP treatment compared with IAQP steel is mainly attributed to the morphology effect of RA. Massive hydrogen-induced cracking (HIC) cracks are found to initiate in the blocky RA of IAQP steel, while only isolate voids are observed in QP steel. It is thus deduced that filmy RA is less susceptible to HE than the blocky RA.

  20. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  1. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-01-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  2. Effect of hydrogen on the behavior of metals II - Hydrogen embrittlement of titanium alloy TV13CA - effect of oxygen - comparison with non-alloyed titanium

    Arditty, Jean-Pierre

    1973-01-01

    The effect of oxygen on the hydrogen embrittlement of non-alloyed titanium and the metastable β titanium alloy, TV13 CA, was studied during dynamic mechanical tests, the concentrations considered varying from 1000 to 5000 ppm (oxygen) and from 0 to 5000 ppm (hydrogen) respectively. TV13 CA alloy has a very high solubility for hydrogen. The establishment of a temperature range and a rate of deformation region in which the embrittlement of the alloy is maximum leads to the conclusion that an embrittlement mechanism occurs involving the dragging and accumulation of hydrogen by dislocations. This is the case for all annealings effected in the medium temperature range, which, by favoring the re-establishment of the stable two-phase α + β state of the alloy, produce hardening. The same is true for oxygen which, in addition to hardening the alloy by the solid solution effect, tends to increase its instability and, in consequence, favors the decomposition of the β phase. Nevertheless oxygen concentrations of up to 1500 ppm contribute to increasing the mechanical resistance without catastrophically reducing the deformation capacity. In the case of non-alloyed titanium, the hardening effect also leads to an increase in E 0.2p c and R, and to a reduction in the deformation capacity. Nevertheless, hydrogen is only very slightly soluble at room temperature and a distribution of the hydride phase linked to the thermal history of the sample predominates. Thus a fine acicular structure obtained from the β phase by quenching, enables an alloy having a good mechanical resistance to be conserved even when large quantities of hydrogen are present; the deformation capacity remains small. On the other hand, when the hydride phase separates the metallic phase into large grains, a very small elongation leads to a breakdown in mechanical resistance. (author) [fr

  3. Radiation embrittlement of WWER-1000 reactor vessel steels

    Nikolaeva, A.V.; Nikolaev, Yu.A.; Kevorkyan, Yu.R.

    2001-01-01

    Results obtained on the blank samples of materials of the WWER-1000 vessels irradiated by low density neutron flux are discussed. Chemical composition of the materials is characterized by the low content of the impurities (copper and phosphorus) and high content of nickel. Dependence of the radiation embrittlement of the WWER-1000 vessel materials on metallurgic variables and damage dose is treated. The research showed that nickel largely enhanced the radiation embrittlement. New dependences for determination of the radiation embrittlement real rate of the WWER-1000 vessel materials and its conservative estimation were developed [ru

  4. Effects of high temperature surface oxides on room temperature aqueous corrosion and environmental embrittlement of iron aluminides

    Buchanan, R.A.; Perrin, R.L.

    1996-09-01

    Studies were conducted to determine the effects of high-temperature surface oxides, produced during thermomechanical processing, heat treatment (750 {degrees}C in air, one hour) or simulated in-service-type oxidation (1000{degrees}C in air, 24 hours) on the room-temperature aqueous-corrosion and environmental-embrittlement characteristics of iron aluminides. Materials evaluated included the Fe{sub 3}Al-based iron aluminides, FA-84, FA-129, FAL and FAL-Mo, a FeAl-based iron aluminide, FA-385, and a disordered low-aluminum Fe-Al alloy, FAPY. Tests were performed in a mild acid-chloride solution to simulate aggressive atmospheric corrosion. Cyclic-anodic-polarization tests were employed to evaluate resistances to localized aqueous corrosion. The high-temperature oxide surfaces consistently produced detrimental results relative to mechanically or chemically cleaned surfaces. Specifically, the pitting corrosion resistances were much lower for the as-processed and 750{degrees} C surfaces, relative to the cleaned surfaces, for FA-84, FA-129, FAL-Mo, FA-385 and FAPY. Furthermore, the pitting corrosion resistances were much lower for the 1000{degrees}C surfaces, relative to cleaned surfaces, for FA-129, FAL and FAL-Mo.

  5. The effects of microstructure on the temper embrittlement susceptibility of a 2 1/4Cr1Mo forging

    Gage, G.; Edwards, B.C.; Hudson, J.A.

    This paper describes the results of a detailed metallurgical assessment of the microstructural stability and temper embrittlement susceptibility of a 255mm thick 2 1/4Cr1Mo steel forging which was manufactured by a process typical of that used for the tube plates of steam generator units. Ageing effects were studied over the temperature range 450-575 deg. C for times up to 20,000h. Grain boundary compositional changes were monitored using Auger Electron Spectroscopy (AES) and microstructural changes determined by both transmission electron microscopy and X-ray analysis. Brittle intergranular failure was produced in the lower shelf energy regime and AES analysis showed that this was associated with the grain boundary segregation of phosphorus. This segregation was shown to exhibit equilibrium characteristics and was consistent with that of phosphorus segregation in α-iron. Implying no significant alloy-impurity interaction. The shift in the ductile-to-brittle transition temperature was not uniquely a function of the grain boundary segregation but was shown to be dependent upon both the level of grain boundary solute segregation and the type of precipitate particles present. Heat treatment conditions which promoted the formation of M 6 C precipitates were particularly deleterious to toughness. (author)

  6. Control of helium effects in irradiated materials based on theory and experiment

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

    1986-01-01

    Helium produced in materials by (n,α) transmutation reactions during neutron irradiations or subjected in ion bombardment experiments causes substantial changes in the response to displacement damage. In particular, swelling, phase transformations and embrittlement are strongly affected. Present understanding of the mechanisms underlying these effects is reviewed. Key theoretical relationships describing helium effects on swelling and helium diffusion are described. Experimental data in the areas of helium effects on swelling and precipitation is reviewed with emphasis on critical experiments that have been designed and evaluated in conjunction with theory. Confirmed principles for alloy design to control irradiation performance are described

  7. Radiation embrittlement behavior of fine-grained molybdenum alloy with 0.2 wt%TiC addition

    Kitsunai, Y. [Tohoku University (Japan); Kurishita, H. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan)]. E-mail: kurishi@imr.tohoku.ac.jp; Kuwabara, T. [Tohoku University (Japan); Narui, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Hasegawa, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Takida, T. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan); Takebe, K. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan)

    2005-11-15

    In order to elucidate the effects of pre-irradiation microstructures and irradiation conditions on radiation embrittlement and radiation-induced ductilization (RIDU), fine-grained Mo-0.2 wt%TiC specimens with high and low reduction rates in plastic working, which are designated as MTC-02H and MTC-02L, respectively, were prepared by powder metallurgical methods. The specimens were neutron irradiated to 0.1-0.15 dpa with controlled 1-cycle and 4-cycle heating between 573 and 773 K, and 473 and 673 K, respectively, in JMTR. Vickers microhardness and three-point bending impact tests and TEM microstructural examinations were made. The degree of radiation embrittlement, assessed by DBTT shift due to irradiation, was strongly dependent on the reduction rate and cycle number. The 4-cycle irradiation suppressed the radiation embrittlement compared with the 1-cycle irradiation, and the suppression was much more significant in MTC-02L than in MTC-02H. The observed behavior is discussed in connection with RIDU and microstructural evolution caused by the 4-cycle irradiation.

  8. Effects of irradiation at low temperature on V-4Cr-4Ti

    Alexander, D.J.; Snead, L.L.; Zinkle, S.J.

    1996-01-01

    Irradiation at low temperatures (100 to 275 degrees C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275 degrees C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation

  9. Effects of irradiation at low temperature on V-4Cr-4Ti

    Alexander, D.J.; Snead, L.L.; Zinkle, S.J. [Oak Ridge National Lab., TN (United States)] [and others

    1996-10-01

    Irradiation at low temperatures (100 to 275{degrees}C) to 0.5 dpa causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile flow properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform ductility. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275{degrees}C. This embrittlement is not due to pickup of O or other interstitial solutes during the irradiation.

  10. Effect of Microstructure and Alloy Chemistry on Hydrogen Embrittlement of Precipitation-Hardened Ni-Based Alloys

    Obasi, G. C.; Zhang, Z.; Sampath, D.; Morana, Roberto; Akid, R.; Preuss, M.

    2018-04-01

    The sensitivity to hydrogen embrittlement (HE) has been studied in respect of precipitation size distributions in two nickel-based superalloys: Alloy 718 (UNS N07718) and Alloy 945X (UNS N09946). Quantitative microstructure analysis was carried out by the combination of scanning and transmission electron microscopy and energy dispersive x-ray spectroscopy (EDS). While Alloy 718 is mainly strengthened by γ″, and therefore readily forms intergranular δ phase, Alloy 945X has been designed to avoid δ formation by reducing Nb levels providing high strength through a combination of γ' and γ″. Slow strain rate tensile tests were carried out for different microstructural conditions in air and after cathodic hydrogen (H) charging. HE sensitivity was determined based on loss of elongation due to the H uptake in comparison to elongation to failure in air. Results showed that both alloys exhibited an elevated sensitivity to HE. Fracture surfaces of the H precharged material showed quasi-cleavage and transgranular cracks in the H-affected region, while ductile failure was observed toward the center of the sample. The crack origins observed on the H precharged samples exhibited quasi-cleavage with slip traces at high magnification. The sensitivity is slightly reduced for Alloy 718, by coarsening γ″ and reducing the overall strength of the alloy. However, on further coarsening of γ″, which promotes continuous decoration of grain boundaries with δ phase, the embrittlement index rose again indicating a change of hydrogen embrittlement mechanism from hydrogen-enhanced local plasticity (HELP) to hydrogen-enhanced decohesion embrittlement (HEDE). In contrast, Alloy 945X displayed a strong correlation between strength, based on precipitation size and embrittlement index, due to the absence of any significant formation of δ phase for the investigated microstructures. For the given test parameters, Alloy 945X did not display any reduced sensitivity to HE compared with

  11. Proton irradiation effects on beryllium – A macroscopic assessment

    Simos, Nikolaos, E-mail: simos@bnl.gov [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Elbakhshwan, Mohamed [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Zhong, Zhong [Photon Sciences, NSLS II, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Camino, Fernando [Center for Functional Nanomaterials, Brookhaven National Laboratory, Upton, NY, 11973 (United States)

    2016-10-15

    Beryllium, due to its excellent neutron multiplication and moderation properties, in conjunction with its good thermal properties, is under consideration for use as plasma facing material in fusion reactors and as a very effective neutron reflector in fission reactors. While it is characterized by unique combination of structural, chemical, atomic number, and neutron absorption cross section it suffers, however, from irradiation generated transmutation gases such as helium and tritium which exhibit low solubility leading to supersaturation of the Be matrix and tend to precipitate into bubbles that coalesce and induce swelling and embrittlement thus degrading the metal and limiting its lifetime. Utilization of beryllium as a pion production low-Z target in high power proton accelerators has been sought both for its low Z and good thermal properties in an effort to mitigate thermos-mechanical shock that is expected to be induced under the multi-MW power demand. To assess irradiation-induced changes in the thermal and mechanical properties of Beryllium, a study focusing on proton irradiation damage effects has been undertaken using 200 MeV protons from the Brookhaven National Laboratory Linac and followed by a multi-faceted post-irradiation analysis that included the thermal and volumetric stability of irradiated beryllium, the stress-strain behavior and its ductility loss as a function of proton fluence and the effects of proton irradiation on the microstructure using synchrotron X-ray diffraction. The mimicking of high temperature irradiation of Beryllium via high temperature annealing schemes has been conducted as part of the post-irradiation study. This paper focuses on the thermal stability and mechanical property changes of the proton irradiated beryllium and presents results of the macroscopic property changes of Beryllium deduced from thermal and mechanical tests.

  12. Oak Ridge National Laboratory Embrittlement Data Base (EDB) and Dosimetry Evaluation (DE) program

    Pace, J.V. III; Remec, I.; Wang, J.A.; White, J.E.

    1996-01-01

    The objective of this program is to develop, maintain, and upgrade computerized data bases, calculational procedures, and standards relating to reactor pressure vessel fluence spectra determinations and embrittlement assessments. As part of this program, the information from radiation embrittlement research on nuclear reactor pressure vessel steels and from power reactor surveillance reports is maintained in a data base published on a periodic basis. The Embrittlement Data Base (EDB) effort consists of verifying the quality of the EDB, providing user-friendly software to access and process the data, and exploring and assessing embrittlement prediction models. The Dosimetry Evaluation effort consists of maintaining and upgrading validated neutron and gamma radiation transport procedures, maintaining cross-section libraries with the latest evaluated nuclear data, and maintaining and updating validated dosimetry procedures and data bases. The information available from this program provides data for assisting the Office of Nuclear Reactor Regulation, with support from the Office of Nuclear Regulatory Research, to effectively monitor current procedures and data bases used by vendors, utilities, and service laboratories in the pressure vessel irradiation surveillance program

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

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

    1998-01-01

    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 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

  14. Updated embrittlement trend curve for reactor pressure vessel steels

    Kirk, M.; Santos, C.; Eason, E.; Wright, J.; Odette, G.R.

    2003-01-01

    The reactor pressure vessels of commercial nuclear power plants are subject to embrittlement due to exposure to high energy neutrons from the core. Irradiation embrittlement of RPV belt-line materials is currently evaluated using US Regulatory Guide 1.99 Revision 2 (RG 1.99 Rev 2), which presents methods for estimating the Charpy transition temperature shift (ΔT30) at 30 ft-lb (41 J) and the drop in Charpy upper shelf energy (ΔUSE). A more recent embrittlement model, based on a broader database and more recent research results, is presented in NUREG/CR-6551. The objective of this paper is to describe the most recent update to the embrittlement model in NUREG/CR-6551, based upon additional data and increased understanding of embrittlement mechanisms. The updated ΔT30 and USE models include fluence, copper, nickel, phosphorous content, and product form; the ΔT30 model also includes coolant temperature, irradiation time (or flux), and a long-time term. The models were developed using multi-variable surface fitting techniques, understanding of the ΔT30 mechanisms, and engineering judgment. The updated ΔT30 model reduces scatter significantly relative to RG 1.99 Rev 2 on the currently available database for plates, forgings, and welds. This updated embrittlement trend curve will form the basis of revision 3 to Regulatory Guide 1.99. (author)

  15. U.S. NRC Embrittlement Data Base (EDB)

    Pace, J.V.; Rosseel, T.M.; Wang, J.A.

    1999-01-01

    Large amounts of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. Version 1 of the Embrittlement Data Base (EDB) [I] is such a comprehensive collection of such data resulting from the merging of the Power Reactor Embrittlement Data Base (PR-EDB) [2] and the Test Reactor Embrittlement Data Base (TR-EDB) [3]. Fracture toughness data were also integrated into Version 1 of the EDB. The EDB data files are in dBASE format and can be accessed with a personal computer using the DOS or WINDOWS operating system. A utility program has been written to investigate radiation embrittlement using this data base. The utility program is used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to tit and plot Charpy impact data

  16. The effect of segregated sp-impurities on grain-boundary and surface structure, magnetism and embrittlement in nickel

    Všianská, Monika; Šob, Mojmír

    2011-01-01

    Roč. 56, č. 6 (2011), s. 817-840 ISSN 0079-6425 R&D Projects: GA AV ČR IAA100100920; GA MŠk(CZ) OC10008; GA ČR GD106/09/H035 Institutional research plan: CEZ:AV0Z20410507 Keywords : grain boundaries * segregation * nickel * embrittlement Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 18.216, year: 2011

  17. Effect of microplastic strain on hydrogen behaviour in steel and resistance to hydrogen embrittlement

    Gribanova, L.I.; Sarrak, V.I.; Filippov, G.A.; Shlyafirner, A.M. (Tsentral' nyj Nauchno-Issledovatel' skij Inst. Chernoj Metallurgii, Moscow (USSR))

    1981-01-01

    A connection between the tendency to delayed fracture and resistance to microplastic deformation is studied in the presence of hydrogen on smooth samples of the 40Kh steel. Tests for delayed fracture have been carried out at the ''Instron'' machine. Two critical levels of strains during delayed fracture in the hydridation process are found out (sigmasub(cr1)=0.3sigmasub(0.2) and sigmasub(cr2)=0.5sigmasub(0.2)). At stresses below the sigmasub(cr1) hydrogen does not influence on the resistance to microplastic deformation of steel and does not cause delayed fracture. Propagation of cracks arising from defects occurring as a result of mutual effect of hydrogen and elastic stresses runs in the stress range from sigmasub(cr1) up to sigmasub(cr2). At stresses higher than sigmasub(cr2) the crack propagates from defects existing in the moment of hydridation process beginning.

  18. Precipitation hardening and hydrogen embrittlement of aluminum ...

    Hydrogen susceptibility of alloy AA7020 was evaluated by slow strain-rate tensile ... high pressures because of the embrittling effect of hydrogen. ... The higher the total Zn + Mg content,. ∗ .... dislocations, leading to a local softening of the slip plane, and thus to ... A Vickers hardness testing machine was used to measure the.

  19. Atmospheric-Induced Stress Corrosion Cracking of Grade 2205 Duplex Stainless Steel—Effects of 475 °C Embrittlement and Process Orientation

    Cem Örnek

    2016-07-01

    Full Text Available The effect of 475 °C embrittlement and microstructure process orientation on atmospheric-induced stress corrosion cracking (AISCC of grade 2205 duplex stainless steel has been investigated. AISCC tests were carried out under salt-laden, chloride-containing deposits, on U-bend samples manufactured in rolling (RD and transverse directions (TD. The occurrence of selective corrosion and stress corrosion cracking was observed, with samples in TD displaying higher propensity towards AISCC. Strains and tensile stresses were observed in both ferrite and austenite, with similar magnitudes in TD, whereas, larger strains and stresses in austenite in RD. The occurrence of 475 °C embrittlement was related to microstructural changes in the ferrite. Exposure to 475 °C heat treatment for 5 to 10 h resulted in better AISCC resistance, with spinodal decomposition believed to enhance the corrosion properties of the ferrite. The austenite was more susceptible to ageing treatments up to 50 h, with the ferrite becoming more susceptible with ageing in excess of 50 h. Increased susceptibility of the ferrite may be related to the formation of additional precipitates, such as R-phase. The implications of heat treatment at 475 °C and the effect of process orientation are discussed in light of microstructure development and propensity to AISCC.

  20. Radiation embrittlement of PWR vessel supports

    Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

    1989-01-01

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

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

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

    1995-01-01

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

  2. Effect of the 718 alloy metallurgical status on hydrogen embrittlement; Effet de l'etat metallurgique de l'alliage 718 sur la fragilisation par l'hydrogene

    Galvano, F.; Andrieu, E.; Blanc, Ch.; Odemer, G.; Ter-Ovanessian, B.; Cocheteau, N.; Holstein, A.; Reboul, Ch. [Universite de Toulouse, CIRIMAT, UPS/CNRS/INPT, 31 - Toulouse (France); Clouez, J.M. [AREVA NP 69 - Lyon (France)

    2010-03-15

    The Inconel 718 is a nickel superalloy which is widely used in the nuclear industry, but is sensitive to hydrogen embrittlement induced by corrosion and stress corrosion cracking phenomena, and by the presence of dissolved hydrogen in pressurized water reactor environments. As this alloy is hardened by precipitation of different intermetallic phases, it appeared that the presence of these precipitates has a strong influence on the hydrogen embrittlement. The authors report the study of the nature and effect of the different traps (intermetallic phases, carbides or their interfaces) on the hydrogen embrittlement susceptibility of the 718 alloy, and more particularly on the observed failure modes. Experiments are performed on tensile samples in which hydrogen content can be measured. The type and grain size of the observed microstructures are given with respect with the thermal treatment, as well as the mechanical properties with or without hydrogen loading

  3. Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

    Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang

    2012-01-01

    The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated

  4. Design and use of the Embrittlement Data Base (EDB)

    Stallmann, F.W.

    1987-01-01

    The architecture of the Embrittlement Data Base (EDB) is described. This data base contains a comprehensive collection of experimental data related to irradiations of reactor pressure vessel steels in surveillance capsules and test reactors. Software is being developed for easy retrieval and analysis of the data. Data and software will be made available to interested parties on a cooperative basis

  5. Reactor pressure vessel embrittlement

    1992-07-01

    Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

  6. SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation, Annealing and Re-Embrittlement

    Van Walle, E.; Chaouadi, R.; Puzzolante, J.L.; Fabry, A.; Van de Velde, J.

    1998-01-01

    The contribution of the Belgian Nuclear Research Centre SCK-CEN to the IAEA Round Robin Exercise on WWER-440 RPV weld material is reported. The objective of this contribution is twofold: (1) to gain experience in the field of the testing of WWER-440 steels; (2) to analyse the round-robin data according to in-house developed on used models in order to check their validity and applicability. Results from testing on unirradiated material are reported including data obtained from chemical analysis, Charpy-V impact testing, tensile testing and fracture toughness determination. Finally, irradiation strategies that can be used in the program to obtain irradiated, irradiated-annealed and irradiated-annealed-reirradiated conditions are outlined

  7. Irradiation Effects Test Series: Test IE-3. Test results report

    Farrar, L.C.; Allison, C.M.; Croucher, D.W.; Ploger, S.A.

    1977-10-01

    The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m 2 . After a flow reduction to 2120 kg/s-m 2 , film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions

  8. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  9. Assessment of thermal aging embrittlement in a cast stainless steel valve and its effect on the structural integrity

    Cicero, S.; Setien, J.; Gorrochategui, I.

    2009-01-01

    This paper analyzes the thermal aging embrittlement occurred in a cast stainless steel valve, which is part of the reactor water clean-up (RWCU) system of a Spanish boiling water reactor (BWR) nuclear power plant. The aim is to estimate the current and future state of the material and the corresponding structural integrity of the valve. Given that there is no data available for the experimental characterization of the material, the evolution of the mechanical properties (fracture toughness, yield stress, flow stress and Ramberg-Osgood parameters) has been estimated using the ANL procedure. With the obtained estimations, the critical crack size has been calculated using the European procedure FITNET FFS and the ASME Code. This analysis considers not only the evolution of the mechanical properties up to now but also its future evolution in case there is a life extension of the plant until year 2029

  10. Effect of γ-IRRADIATION on the Mechanical Properties of Al-Cu Alloy

    Abo-Elsoud, M.; Ismail, H.; Sobhy, Maged S.

    SEM observations and Vickers hardness tests were performed to identify the irradiation effects. γ-irradiation effect during the aging hardening process can be explained depending on the composition of the alloy and is used to derive quantitative information on the kinetics of the transformation precipitates. Increasing the Cu content of an Al-Cu alloy can improve the aging hardness. The present results of the hardness behavior, with SEM observations of surveillance specimens at different doses, suggest that the radiation-induced defects are probably complex valence-solute clusters. These clusters act as nuclei for the precipitation of θ-Al2Cu type. This can be effectively utilized to study the systematics of nucleation of precipitates at vacancy-type defects. γ-irradiation probably plays the key role in defects responsible for material strengthening and embrittlement.

  11. Postirradiation examination results for the Irradiation Effects Test IE-5

    Cook, T.F.; Ploger, S.A.; Hobbins, R.R.

    1978-03-01

    The results are presented of the postirradiation examination of four pressurized water reactor type fuel rods which were tested in-pile under a fast power ramp and film boiling operation during Irradiation Effects (IE) Test 5. The major objectives of this test were to evaluate the effects of simulated fission products on fuel rod behavior during a fast power ramp, to determine the effects of high initial internal pressure on a fuel rod during film boiling, and to assess fuel rod property changes that occur during film boiling in a fuel rod with previously irradiated cladding. The overall condition of the rods and changes that occurred in fuel and cladding as a result of the power ramp and film boiling operation, as determined from the postirradiation examination, are reported and analyzed. Effects of the simulated fission products on fuel rod behavior during a power ramp are discussed. The effect of high internal pressure on rod behavior during film boiling is evaluated. Cladding temperatures are estimated at various axial and circumferential locations. Cladding embrittlement by oxidation is also assessed

  12. Different approaches to estimation of reactor pressure vessel material embrittlement

    V. M. Revka

    2013-03-01

    Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

  13. Nanocrystalline Steels’ Resistance to Hydrogen Embrittlement

    Skołek E.

    2015-04-01

    Full Text Available The aim of this study is to determine the susceptibility to hydrogen embrittlement in X37CrMoV5-1 steel with two different microstructures: a nanocrystalline carbide-free bainite and tempered martensite. The nanobainitic structure was obtained by austempering at the bainitic transformation zone. It was found, that after hydrogen charging, both kinds of microstructure exhibit increased yield strength and strong decrease in ductility. It has been however shown that the resistance to hydrogen embrittlement of X37CrMoV5-1 steel with nanobainitic structure is higher as compared to the tempered martensite. After hydrogen charging the ductility of austempered steel is slightly higher than in case of quenched and tempered (Q&T steel. This effect was interpreted as a result of phase composition formed after different heat treatments.

  14. Mechanisms of liquid-metal embrittlement

    Popovich, V.V.

    1979-01-01

    The mechanism of the embrittlement of metals and alloys during deformation in contact with liquid metals are discussed. With 20Kh13 steel in a Pb-Sn melt and polycrystalline Al in the presence of various mercury solutions a.s examples, considered are the three main processes - adsorption, corrosion (dissolution), formation of new phases which cause the disintegration of materials under the action of liquid-metallic media. Presented are data on plastic ductile and strength properties of the above materials in the presence of liquid-metallic media. A model is described that takes into account the effect of the medium upon the plastic deformation and the part the medium plays in liquid-metallic embrittlement

  15. Estimation of RPV material embrittlement for Ukrainian NPP based on surveillance test data

    Revka, V.; Chyrko, L.; Chaikovsky, Yu.; Gulchuk, Yu.

    2012-01-01

    The WWER-1000 RPV material embrittlement has been evaluated using the surveillance test data for the nuclear power plant which is under operation in Ukraine. The RPV materials after the neutron (E > 0,5 MeV) irradiation up to fluence of 22,9.10 22 m -2 have been studied. Fracture toughness tests were performed using pre-cracked Charpy specimens for the beltline materials (base and weld metal). The maximum shift of T 0 reference temperature is equal to 44 o C. A radiation embrittlement rate, A F , for the RPV materials was estimated using the standard and reconstituted specimens. A comparison of the A F values has shown a good agreement between the specimen sets before and after reconstitution both for base and weld metal. Furthermore it has been revealed there is no nickel effect for the studied materials. In spite of the high nickel content the radiation embrittlement rate for weld metal is not higher than for base metal with low nickel content. Fracture toughness analysis has shown the Master curve shape describes well a temperature dependence of K Jc values. However a higher scatter of K Jc values is observed in comparison to 95 % tolerance bounds. (author)

  16. Simulation of He embrittlement at grain boundaries in bcc transition metals

    Suzudo, Tomoaki; Yamaguchi, Masatake

    2015-01-01

    To investigate what atomic properties largely determine vulnerability to He embrittlement at grain boundaries (GB) of bcc metals, we introduce a computational model composed of first principles density functional theory and a He segregation rate theory model. Predictive calculations of He embrittlement at the first wall of the future DEMO fusion concept reactor indicate that variation in the He embrittlement originated not only from He production rate related to neutron irradiation, but also from the He segregation energy at the GB that has a systematic trend in the periodic table. - Highlights: • We modeled He grain boundary (GB) segregation of bcc transition metals using first-principles-based rate theory. • We established the quantitative relation between He embrittlement and He segregation using GB cohesive energy. • He embrittlement was strongly dependent on He segregation energy at the GB that has a systematic trend in the periodic table.

  17. Simulation of He embrittlement at grain boundaries in bcc transition metals

    Suzudo, Tomoaki, E-mail: suzudo.tomoaki@jaea.go.jp; Yamaguchi, Masatake

    2015-10-15

    To investigate what atomic properties largely determine vulnerability to He embrittlement at grain boundaries (GB) of bcc metals, we introduce a computational model composed of first principles density functional theory and a He segregation rate theory model. Predictive calculations of He embrittlement at the first wall of the future DEMO fusion concept reactor indicate that variation in the He embrittlement originated not only from He production rate related to neutron irradiation, but also from the He segregation energy at the GB that has a systematic trend in the periodic table. - Highlights: • We modeled He grain boundary (GB) segregation of bcc transition metals using first-principles-based rate theory. • We established the quantitative relation between He embrittlement and He segregation using GB cohesive energy. • He embrittlement was strongly dependent on He segregation energy at the GB that has a systematic trend in the periodic table.

  18. Effects of 1000 C oxide surfaces on room temperature aqueous corrosion and environmental embrittlement of iron aluminides

    Buchanan, R.A.; Perrin, R.L. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering

    1997-12-01

    Results of electrochemical aqueous-corrosion studies at room temperature indicate that retained in-service-type high-temperature surface oxides (1000 C in air for 24 hours) on FA-129, FAL and FAL-Mo iron aluminides cause major reductions in pitting corrosion resistance in a mild acid-chloride solution designed to simulate aggressive atmospheric corrosion. Removal of the oxides by mechanical grinding restores the corrosion resistance. In a more aggressive sodium tetrathionate solution, designed to simulate an aqueous environment contaminated by sulfur-bearing combustion products, only active corrosion occurs for both the 1000 C oxide and mechanically cleaned surfaces at FAL. Results of slow-strain-rate stress-corrosion-cracking tests on FA-129, FAL and FAL-Mo at free-corrosion and hydrogen-charging potentials in the mild acid chloride solution indicate somewhat higher ductilities (on the order of 50%) for the 1000 C oxides retard the penetration of hydrogen into the metal substrates and, consequently, are beneficial in terms of improving resistance to environmental embrittlement. In the aggressive sodium tetrathionate solution, no differences are observed in the ductilities produced by the 1000 C oxide and mechanically cleaned surfaces for FAL.

  19. Irradiation effects on perfluorinated polymers

    Lappan, U.; Geissler, U.; Haeussler, L.; Pompe, G.; Scheler, U.; Lunkwitz, K.

    2002-01-01

    Complete text of publication follows. High-energy radiation affects the properties of polymers by chain scission and crosslinking reactions. Both types of reaction occur simultaneously in irradiated polymers. However, one process will usually predominate, depending on the chemical structure of the polymer and the irradiation conditions such as temperature and atmosphere. Polytetrafluoroethylene (PTFE) undergoes predominantly chain scission, if the irradiation is performed at room temperature. This shortcoming is exploited by converting PTFE into low molecular weight micropowders. The use of PTFE micropowders functionalized with COOH groups as additive in polyamides to improve the sliding properties of the materials has been studied. During the compounding process in a twin screw extruder the COOH groups of the irradiated PTFE react with the polyamides. For these studies, it became necessary to investigate the content of end groups in irradiated PTFE by FTIR and 19 F solid-state NMR. These date were used to calculate number-average molecular weights. The ratios of COOH groups to CF 3 groups are discussed in terms of the mechanism of PTFE degradation. If PTFE is irradiated at temperatures above its crystalline melting point in an oxygen-free atmosphere, branching and crosslinking occur. The dependence of radiation effects on perfluorinated copolymers (FEP, PFA) on temperature has been studied. Melt flow index measurements have shown that branching and crosslinking predominate over chain scission with increasing irradiation temperature both in FEP and in PFA. Quantitative analysis of 19 F solid-state NMR data has shown that the content of branching groups (>CF-) exceeds the content of end groups in the case of PFA irradiated above its crystalline melting point. The formation of COF and COOH groups in the irradiated PFA is interpreted as a result of partial degradation of perfluorovinyl ether comonomer units

  20. Experimental data base for assessment of irradiation induced ageing effects in pre-irradiated RPV materials of German PWR

    Hein, H.; Gundermann, A.; Keim, E.; Schnabel, H. [AREVA NP GmbH (Germany); Ganswind, J. [VGB PowerTech e.V (Germany)

    2011-07-01

    The 5 year research program CARISMA which ended in 2008 has produced a data base to characterize the fracture toughness of pre-irradiated original RPV (Reactor Pressure Vessel) materials being representative for all four German PWR construction lines of former Siemens/KWU company. For this purpose tensile, Charpy-V impact, crack initiation and crack arrest tests have been performed for three base materials and four weld metals irradiated to neutron fluences beyond the designed EoL range. RPV steels with optimized chemical composition and with high copper as well as high nickel content were examined in this study. The RTNDT concept and the Master Curve approach were applied for the assessment of the generated data in order to compare both approaches. A further objective was to clarify in which extent crack arrest curves can be generated for irradiated materials and how crack arrest can be integrated into the Master Curve approach. By the ongoing follow-up project CARINA the experimental data base will be extended by additional representative materials irradiated under different conditions and with respect to the accumulated neutron fluences and specific impact parameters such as neutron flux and manufacturing effects. The irradiation data cover also the long term irradiation behavior of the RPV steels concerned. Moreover, most of the irradiated materials were and will be used for microstructural examinations to get a deeper insight in the irradiation embrittlement mechanisms and their causal relationship to the material property changes. By evaluation of the data base the applicability of the Master Curve approach for both crack initiation and arrest was confirmed to a large extent. Moreover, within both research programs progress was made in the development of crack arrest test techniques and in specific issues of RPV integrity assessment. (authors)

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

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

    1986-01-01

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

  2. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  3. Irradiation effects test Series Scoping Test 1: test results report

    Quapp, W.J.; Allison, C.M.; Farrar, L.C.

    1977-09-01

    The report describes the results of the first scoping test in the Irradiation Effects Test Series conducted by the Thermal Fuels Behavior Program, which is part of the Water Reactor Research Program of EG and G Idaho, Inc. The research is sponsored by the United States Nuclear Regulatory Commission. This test used an unirradiated, three-foot-long, PWR-type fuel rod. The objective of this test was to thoroughly evaluate the remote fabrication procedures to be used for irradiated rods in future tests, handling plans, and reactor operations. Additionally, selected fuel behavior data were obtained. The fuel rod was subjected to a series of preconditioning power cycles followed by a power increase which brought the fuel rod power to about 20.4 kW/ft peak linear heat rating at a coolant mass flux of 1.83 x 10 6 lb/hr-ft 2 . Film boiling occurred for a period of 4.8 minutes following flow reductions to 9.6 x 10 5 and 7.5 x 10 5 lb/hr-ft 2 . The test fuel rod failed following reactor shutdown as a result of heavy internal and external cladding oxidation and embrittlement which occurred during the film boiling operation

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

    Gaganidze, E., E-mail: Ermile.Gaganidze@kit.edu [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiss, O.J.; Aktaa, J. [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V. [Joint Stock Company ' State Scientific Centre Research Institute of Atomic Reactors' , 433510 Dimitrovgrad-10, Ulyanovsk Region (Russian Federation)

    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 deg. 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.

  5. Intrinsic ductility and environmental embrittlement of binary Ni3Al

    George, E.P.; Liu, C.T.; Pope, D.P.

    1993-01-01

    Polycrystalline, B-free Ni 3 Al (23.4 at.% Al), produced by cold working and recrystallizing a single crystal, exhibits room temperature tensile ductilities of 3-5% in air and 13-16% in oxygen. These ductilities are considerably higher than anything previously reported, and demonstrate that the 'intrinsic' ductility of Ni 3 Al is much higher than previously thought. They also show that the moisture present in ordinary ambient air can severely embrittle Ni 3 Al (ductility decreasing from a high of 16% in oxygen to a low of 3% in air). Fracture is predominantly intergranular in both air and oxygen. This indicates that, while moisture can further embrittle the GBs in Ni 3 Al, they persist as weak links even in the absence of environmental embrittlement. However, they are not 'intrinsically brittle' as once thought, since they can withstand relatively large plastic deformations prior to fracture. Because B essentially eliminates environmental embrittlement in Ni 3 Al - and environmental embrittlement is a major cause of poor ductility in B-free Ni 3 Al - it is concluded that a significant portion of the so-called B effect must be related to suppression of moisture-induced environmental embrittlement. However, since B-doped Ni 3 Al fractures transgranularly, whereas B-free Ni 3 Al fractures predominantly intergranularly, B must have the added effect that it strengthens the GBs. A comparison with the earlier work on Zr-doped Ni 3 Al shows that Zr improves the ductility of Ni 3 Al, both in air and (and even more dramatically) in oxygen. While the exact mechanism of this ductility improvement is not clear at present, Zr appears to have more of an effect on (enhancing) GB strength than on (suppressing) environmental embrittlement

  6. Liquid and Solid Metal Embrittlement.

    1981-09-05

    example, embrittlement of AISI 4140 steel begins at T/T, - 0.75 for cadmium, and 0.85 for lead and tin environments (2). In a few cases, e.g. zinc...has recently proposed, however, that liquid zinc can penetrate to very near the tip of a sharp crack in 4140 steel, based upon both direct observation...long could be detected, was observed in delayed failure experi- ments on unnotched 4140 steel, in the quenched and tempered condi- tion, embrittled by

  7. Reactor pressure vessel embrittlement management through EPRI-Developed material property databases

    Rosinski, S.T.; Server, W.L.; Griesbach, T.J.

    1997-01-01

    Uncertainties and variability in U.S. reactor pressure vessel (RPV) material properties have caused the U.S. Nuclear Regulatory Commission (NRC) to request information from all nuclear utilities in order to assess the impact of these data scatter and uncertainties on compliance with existing regulatory criteria. Resolving the vessel material uncertainty issues requires compiling all available data into a single integrated database to develop a better understanding of irradiated material property behavior. EPRI has developed two comprehensive databases for utility implementation to compile and evaluate available material property and surveillance data. RPVDATA is a comprehensive reactor vessel materials database and data management program that combines data from many different sources into one common database. Searches of the data can be easily performed to identify plants with similar materials, sort through measured test results, compare the ''best-estimates'' for reported chemistries with licensing basis values, quantify variability in measured weld qualification and test data, identify relevant surveillance results for characterizing embrittlement trends, and resolve uncertainties in vessel material properties. PREP4 has been developed to assist utilities in evaluating existing unirradiated and irradiated data for plant surveillance materials; PREP4 evaluations can be used to assess the accuracy of new trend curve predictions. In addition, searches of the data can be easily performed to identify available Charpy shift and upper shelf data, review surveillance material chemistry and fabrication information, review general capsule irradiation information, and identify applicable source reference information. In support of utility evaluations to consider thermal annealing as a viable embrittlement management option, EPRI is also developing a database to evaluate material response to thermal annealing. Efforts are underway to develop an irradiation

  8. Status on the selection and development of an embrittlement trend curve to use in ASTM standard guide E900

    Kirk, M.; Brian Hall, J.; Server, W.; Lucon, E.; Erickson, M.; Stoller, R.

    2015-01-01

    ASTM E900-07, Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, includes an embrittlement trend curve. The trend curve can be used to predict the effect of neutron irradiation on the embrittlement of ferritic pressure vessel steels, as quantified by the shift in the Charpy V-Notch transition curve at 41 Joules of absorbed energy (ΔT 41J ). The current E900 trend curve was first adopted in the 2002 revision. In 2011 ASTM Subcommittee E10.02 undertook an extensive effort to evaluate the adequacy of the E900 trend curve for continued use. This paper summarizes the current status of this effort, which has produced a trend curve calibrated using a database of over 1800 ΔT 41J values from the light water reactor surveillance programs in thirteen countries. (authors)

  9. Neutron irradiation effect on the strength of jointed Ti-6Al-4V alloy

    Ishiyama, Shintaro; Miya, Naoyuki

    2002-01-01

    In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8 x 10 23 n/m 2 (>0.18 MeV). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of fracture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation. (author)

  10. Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

    Krasikov, E. A.

    2012-01-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated

  11. Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

    Krasikov, E.A.

    2012-01-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 o C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 o C and following extra irradiation (87 h at 330 o C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated management methods, will help

  12. Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

    Krasikov, E. A. [National Research Centre Kurchatov Inst., 1, Kurchatov Sq., Moscow, 123182 (Russian Federation)

    2012-07-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which

  13. Effects of irradiation upon spices

    1978-04-01

    ESR studies were performed on untreated and irradiated samples of paprika powder, ground black pepper, and a spice mixture of the following composition: paprika, 55%; black pepper, 14%; allspice, 9%; coriander, 9%; marjoram, 7%; cumin, 4%; and nutmeg, 2%. Gamma radiation doses from 0.5 to 5 Mrad were applied. In the case of paprika samples, the effect of moisture content on the formation and disappearance of radiation-induced free radicals was also investigated. Shortly after irradiation (on the day of radiation treatment) high amounts of free radicals were detected in irradiated spice samples but they diminished upon storage. After a period of 3 months the ESR signals of the irradiated samples approximated those of the controls. The free radicals found in unirradiated ground spices did not disappear during a storage period as long as one year. The formation and disappearance of radiation-induced free radicals were found to be strongly affected by the moisture content of samples. If a sample of low moisture content containing a high free radical concentration after irradiation was placed in an atmosphere of higher moisture content, the free radicals decayed rapidly.

  14. Power reactor embrittlement data base

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1989-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs

  15. Neutron irradiation effects on grain-refined W and W-alloys

    Hasegawa, A.; Fukuda, M.; Tanno, T.; Nogami, S.; Yabuuchi, K.; Tanaka, T.; Muroga, T.

    2014-10-01

    Microstructural data of neutron irradiated Tungsten (W) such as size and number density of voids and precipitates obtained by W up to 1.5dpa irradiation in the temperature range of 400-800degC were compiled quantitatively. Nucleation and growth process of these defects were clarified and a qualitative prediction of the damage structure development and hardening of W in fusion reactor environments were made taking into account the solid transmutation effects for the first time. To improve recrystallization behavior and low temperature embrittlement, grain refined-W alloys were fabricated by K- or La-doping method. Rhenium addition to the grain refining process was also examined to improve mechanical properties. Characterizations of unirradiated materials were performed. (author)

  16. Progress on untargeted effects of ionizing irradiation

    Liu Jing; Chen Jihong; Li Wenjian

    2010-01-01

    The side effect of ionizing irradiation has been paid more attention with its widely using in tumor treating and mutation breeding. In recent years, untargeted effects induced by ionizing irradiation have become a hotspot of radiobiology. Here, according to reported results, we reviewed the types (genomic instability, bystander effect and adaptive response) and mechanisms of untargeted effects of ionizing irradiation in this paper. (authors)

  17. Low temperature hydrogen embrittlement of niobium. II. Microscopic observations

    Grossbeck, M.L.; Birnbaum, H.K.

    1977-01-01

    The detailed, microscopic processes which occur during the hydrogen embrittlement of pure Nb are examined using in situ SEM crack propagation studies, SEM fractography, electron diffraction and ion probe methods. These results show that the fracture process occurs in a stress induced NbH hydride phase which forms in front of the propagating crack. The experimental results are in good agreement with the stress induced hydride embrittlement mechanism which is discussed. The thermodynamics of precipitation of hydrides under external stress is discussed and calculations are presented for the stress effects on the α-β solvus temperatures. These are related to the embrittlement process and evidence is presented to support the calculated stress effects on the solvus temperature

  18. Mercury embrittlement of Cu-Al alloys under cyclic loading

    Regan, T. M.; Stoloff, N. S.

    1977-01-01

    The effect of mercury on the room temperature, high cycle fatigue properties of three alloys: Cu-5.5 pct Al, Cu-7.3 pct Al, and Cu-6.3 pct Al-2.5 pct Fe has been determined. Severe embrittlement under cyclic loading in mercury is associated with rapid crack propagation in the presence of the liquid metal. A pronounced grain size effect is noted under mercury, while fatigue properties in air are insensitive to grain size. The fatigue results are discussed in relation to theories of adsorption-induced liquid metal embrittlement.

  19. PR-EDB: Power Reactor Embrittlement Database Version 3

    Wang, Jy-An John; Subramani, Ranjit

    2008-01-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  20. PR-EDB: Power Reactor Embrittlement Database - Version 3

    Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  1. Grain boundary embrittlement and cohesion enhancement in copper

    Paxton, Anthony; Lozovoi, Alexander [Atomistic Simulation Centre, Queen' s University Belfast, BT7 1NN (United Kingdom); Schweinfest, Rainer [Science+Computing ag, Hagellocher Weg 71-5, 720270 T ubingen (Germany); Finnis, Michael [Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom)

    2008-07-01

    There has been a long standing debate surrounding the mechanism of grain boundary embrittlement and cohesion enhancement in metals. Embrittlement can lead to catastrophic failure such as happened in the Hinkley Point disaster, or indeed in the case of the Titanic. This kind of embrittlement is caused by segregation of low solubility impurities to grain boundaries. While the accepted wisdom is that this is a phenomenon driven by electronic or chemical factors, using language such as charge transfer and electronegativity difference; we believe that in copper, at least, both cohesion enhancement and reduction are caused by a simple size effect. We have developed a theory that allows us to separate unambiguously, if not uniquely, chemical and structural factors. We have studied a large number of solutes in copper using first principles atomistic simulation to support this argument, and the results of these calculations are presented here.

  2. Power Reactor Embrittlement Data Base

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1990-01-01

    Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

  3. Effect of niobium on the embrittlement of 2.25 Cr and 2.25 Cr-1Mo steels by phosphous

    Antunes, J.L.B.

    1985-01-01

    The influence of niobium on the temper embrittlement of 2.25Cr and 2.25 Cr-1Mo steels doped with phosphorus is evaluated. The transition temperatures of the samples tempered at 650 0 C and aged at different temperatures for niobium steels. (M.J.C.) [pt

  4. Low dose irradiation effects on DIN 1.4948 mechanical properties

    Schaaf, B. van der; Vries, M.I. de

    For the SNR 300 the licensing authorities require the determination of the lower boundaries of post-irradiation mechanical properties for DIN 1.4948 parent metal and welded joints. It has been established that with decreasing strain rate the post-irradiation tensile ductility decreases. A transition strain rate has been observed, above which there is no effect of irradiation on ductility. The transition strain rate increases with increasing temperature. Coarse grained heats show lower ultimate tensile strength above 800 K than fine grained heats. There is no significant effect of irradiation on load controlled high cycle fatigue with frequencies of 1 Hz or higher. In low cycle fatigue numbers of cycles to failure decrease with decreasing frequency. Increasing the test temperature reduces the number of cycles to failure even more. The frequency effect is more evident at 823 K. Parent metal has a better fatigue resistance than welded joints in unirradiated and irradiated condition. Creep strength is reduced by irradiation due to loss of ductility. It is shown that with increasing grain size the rupture strength decreases. The ductility of welded joints after irradiation is low, in some cases as low as 0.5% creep strain. After irradiation, tensile, creep and fatigue fracture surfaces show many more intergranular features than in the equivalent unirradiated condition. The promotion of intergranular fracture by irradiation and the consequent degradation of low strain rate mechanical properties is explained by the presence of helium on grain boundaries. Several measures to increase the helium content threshold can be taken, such as grain refinement, homogeneous boron distribution and promotion of helium bubble initiation. In cases where helium embrittlement is encountered, life reduction factors on unirradiated material properties must be applied

  5. Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

    Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

  6. Fracture analysis of HFIR beam tube caused by radiation embrittlement

    Chang, S.J.

    1994-01-01

    With an attempt to estimate the neutron beam tube embrittlement condition for the Oak Ridge High Flux Isotope Reactor (HFIR), fracture mechanics calculations are carried out in this paper. The analysis provides some numerical result on how the tube has been structurally weakened. In this calculation, a lateral impact force is assumed. Numerical result is obtained on how much the critical crack size should be reduced if the beam tube has been subjected to an extended period of irradiation. It is also calculated that buckling strength of the tube is increased, not decreased, with irradiation

  7. Irradiation induced effects in zirconium (A review)

    Madden, P.K.

    1975-06-01

    Irradiation creep in zirconium and its alloys is comprehensively discussed. The main theories are outlined and the gaps between them and the observed creep behaviour, indicated. Although irradiation induced point defects play an important role, effects due to irradiation induced dislocation loops seem insignificant. The experimental results suggest that microstructural variations due to prior cold-working or hydrogen injection perturb the irradiation growth and the irradiation creep of zircaloy. Further investigations into these areas are required. One disadvantage of creep experiments lies in their duration. The possibility of accelerated experiments using ion implantation or electron irradiation is examined in the final section, and its possible advantages and disadvantages are outlined. (author)

  8. Thermal annealing of an embrittled reactor pressure vessel

    Mager, T.R.; Dragunov, Y.G.; Leitz, C.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

  9. Thermal embrittlement of reactor vessel steels

    Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

    1995-01-01

    As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

  10. Effect of irradiation on sweet corn preservation

    Fu Junjie

    2002-01-01

    60 Co γ-ray was used to irradiate newly-harvested sweet corn and the results showed that the effects of irradiation on soluble solids, sucrose, starch and total sugar were not significant. The viscosity of starch decreased with the increasing of irradiation dose. The preservation duration of irradiated sweet corn was 7 days longer than that of CK, and the sweet, smell, taste of sweet corn had no abnormal change

  11. Neutron irradiation effects on plasma facing materials

    Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

    2000-12-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

  12. Neutron irradiation effects on plasma facing materials

    Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

    2000-01-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

  13. Irradiation effects on mechanical properties of fuel element cladding from thermal reactors

    Chatterjee, S.

    2005-01-01

    During reactor operation, UO 2 expands more than the cladding tube (Zirconium alloys for thermal reactors), is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount (if) deposited on the cladding, and hydride morphology (e.g. hydride lenses). The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings

  14. Effect of irradiation on carbohydrates content

    Chantharasakul, S.

    1971-01-01

    Effect of gamma radiation on vitamin C and total acidity contents of Hom Tong banana was described. There was a slight decrease in vitamin C contents in both irradiated and non-irradiated banana during storage. No difference was detected in term of vitamin C contents between irradiated and non-irradiated banana at any storage time. The total acidity of the banana increased with increasing time of storage owing to the ripening effect of the fruit. Higher total acidity content of non-irradiated banana during storage indicated the faster rate of ripening of the fruit

  15. Estimation of embrittlement damage risk at neutron embrittled vessel constructions

    Staevski, K.; Madzharov, D.; Detistov, P.; Petrova, T.

    1998-01-01

    In this work a methodology based on Damage mechanics criteria is proposed. This methodology serves for probability assessment of the brittle damage risk for the neutron embrittled vessel elements. The developed methodology is realised in RISK code and has been verified on the base of tough reliability of the pressure vessel, 'Kozloduy' NPP Unit 2. This investigation has been carried out at the given parameters of the possible defects on the vessel's weld 4 taking into account requirements of the western and Russian standards. The obtained values for ductile to brittle transition temperatures, defining the equipment life-time in the presence of maximal defect, are in good consistence with the experimentally determined ones. The analyses of results show that the pressure vessel of 'Kozloduy' NPP Unit 2 has got a high level of reliability from brittle damage risk point of view and that the western standards give more conservative evaluation. On the bases of the results a conclusion is made that the developed methodology enables analysing the influence of possible defects in the neutron embrittled elements on their to reliability and their remained life-time

  16. Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

    Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

    2011-01-01

    Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter

  17. Multiscale modelling and experimentation of hydrogen embrittlement in aerospace materials

    Jothi, Sathiskumar

    Pulse plated nickel and nickel based superalloys have been used extensively in the Ariane 5 space launcher engines. Large structural Ariane 5 space launcher engine components such as combustion chambers with complex microstructures have usually been manufactured using electrodeposited nickel with advanced pulse plating techniques with smaller parts made of nickel based superalloys joined or welded to the structure to fabricate Ariane 5 space launcher engines. One of the major challenges in manufacturing these space launcher components using newly developed materials is a fundamental understanding of how different materials and microstructures react with hydrogen during welding which can lead to hydrogen induced cracking. The main objective of this research has been to examine and interpret the effects of microstructure on hydrogen diffusion and hydrogen embrittlement in (i) nickel based superalloy 718, (ii) established and (iii) newly developed grades of pulse plated nickel used in the Ariane 5 space launcher engine combustion chamber. Also, the effect of microstructures on hydrogen induced hot and cold cracking and weldability of three different grades of pulse plated nickel were investigated. Multiscale modelling and experimental methods have been used throughout. The effect of microstructure on hydrogen embrittlement was explored using an original multiscale numerical model (exploiting synthetic and real microstructures) and a wide range of material characterization techniques including scanning electron microscopy, 2D and 3D electron back scattering diffraction, in-situ and ex-situ hydrogen charged slow strain rate tests, thermal spectroscopy analysis and the Varestraint weldability test. This research shows that combined multiscale modelling and experimentation is required for a fundamental understanding of microstructural effects in hydrogen embrittlement in these materials. Methods to control the susceptibility to hydrogen induced hot and cold cracking and

  18. Specificity in liquid metal induced embrittlement

    Fernandes, PJL

    1996-12-01

    Full Text Available One of the most intriguing features of liquid metal induced embrittlement (LMIE) is the observation that some liquid metal-solid metal couples are susceptible to embrittlement, while others appear to be immune. This is referred to as the specificity...

  19. Biological effects of prenatal irradiation

    Streffer, Christian

    1997-01-01

    After large releases of radionuclides, exposure of the embryo or fetus can take place by external irradiation or uptake of radionuclies. The embryo and fetus are radiosensitive throughout prenatal development. The quality and extent of radiation effects depend on the development stage. During the preimplantation period (one to 10 days postconception, p.c.) a radiation exposure of at least 0.2 Gy can cause the death of the embryo. Malformations are only observed in rare cases when genetic predisposition exist. Macroscopic, anatomical malformations are induced only after irradiation during the major organogenesis (two to eight weeks p.c.). A radiation dose of about 0.2 Gy is a doubling dose for the malformation risks as extrapolated from experiments with rodents. The human embryo may be more radioresistant. During early fetogenesis (8-15 weeks p.c.) a high radiosensitivity exists for the developmental of the brain. Radiation doses of 1.0 Gy cause severe mental retardation in about 40% of the exposed fetuses. It must be taken into account that a radiation exposure during the fetal period can also induce cancer. It is generally assumed that the risk exists at about the same level as for children. (Author)

  20. Hydrogen embrittlement of steels: study and prevention

    Brass, A.M.; Chene, J.; Coudreuse, L.

    2000-01-01

    Hydrogen embrittlement of steels is one of the important reason of rupture of pieces in the industry (nuclear, of petroleum..). Indeed, there are a lot of situations which can lead to the phenomenon of hydrogen embrittlement: introduction of hydrogen in the material during the elaboration or during transformation or implementation processes (heat treatments, welding); use of steels when hydrogen or hydrogenated gaseous mixtures are present; hydrogen produced by electrolytic reactions (surface treatments, cathodic protection). The hydrogen embrittlement can appear in different forms which depend of a lot of parameters: material (state, composition, microstructure..); surrounding medium (gas, aqueous medium, temperature..); condition of mechanical solicitation (static, dynamic, cyclic..). The industrial phenomena which appear during cases of hydrogen embrittlement are more particularly described here. Several methods of steels studies are proposed as well as some possible ways for the prevention of hydrogen embrittlement risks. (O.M.)

  1. Embrittlement recovery due to annealing of reactor pressure vessel steels

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1998-01-01

    The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes. (orig.)

  2. The impact of mobile point defect clusters in a kinetic model of pressure vessel embrittlement

    Stoller, R.E.

    1998-05-01

    The results of recent molecular dynamics simulations of displacement cascades in iron indicate that small interstitial clusters may have a very low activation energy for migration, and that their migration is 1-dimensional, rather than 3-dimensional. The mobility of these clusters can have a significant impact on the predictions of radiation damage models, particularly at the relatively low temperatures typical of commercial, light water reactor pressure vessels (RPV) and other out-of-core components. A previously-developed kinetic model used to investigate RPV embrittlement has been modified to permit an evaluation of the mobile interstitial clusters. Sink strengths appropriate to both 1- and 3-dimensional motion of the clusters were evaluated. High cluster mobility leads to a reduction in the amount of predicted embrittlement due to interstitial clusters since they are lost to sinks rather than building up in the microstructure. The sensitivity of the predictions to displacement rate also increases. The magnitude of this effect is somewhat reduced if the migration is 1-dimensional since the corresponding sink strengths are lower than those for 3-dimensional diffusion. The cluster mobility can also affect the evolution of copper-rich precipitates in the model since the radiation-enhanced diffusion coefficient increases due to the lower interstitial cluster sink strength. The overall impact of the modifications to the model is discussed in terms of the major irradiation variables and material parameter uncertainties

  3. Modelling irradiation effect of EUROFER

    Boutard, J.-L.; Dudarev, S.; Victoria, M.

    2006-01-01

    In fusion power reactor, the properties that controlled the behaviour of materials are affected at the atom scale: (i) the crystalline structure is locally destroyed where a displacement cascade occurs, (ii) the chemical bond is affected by transmutation products such as He and H, (iii) an radiation induced microstructure will take place due the diffusion of these point defects and impurities EFDA has launched a programme since 2002 to develop and validate modelling tools to predict the radiation effects in the reference ferritic martensitic steel Eurofer. Up to now, the effort has been devoted (i) to validate the multi-scale modelling approach based on ab-initio energetics map of point defects and He, (iii) to develop inter-atomic potentials for Molecular Dynamics simulation of displacement cascades and dislocation dynamics. Formation and migration energies and diffusion mechanisms of small vacancy (n< ) and interstitial clusters (n< ) were computed with the ab-initio code SIESTA and used to successfully predict via Kinetic Monte Carlo the experimental recovery stages of radiation damage in ultra high purity Fe. A complete He and point defect energetics mapping was ab-initio determined in Fe-C and used to reproduce via Rate Theory He-desorption from pre-implanted specimens. A developed '' magnetic '' potential is capable of transferring the magnetic properties of Fe due to the 3d-electron correlation to the scale of the Molecular Dynamics. An inter-atomic potential is being developed to reproduce the thermodynamics of the Fe-Cr system. The program will now be devoted (i) to develop atom-scale reference kinetic methods to predict the phase - stability of the Fe-Cr thermally and under irradiation (ii) to predict at the atom scale the core structure and dynamics of screw dislocation and their collective behaviour at the meso-scale, using Discrete Dislocation Dynamics (iii) to validate at the relevant scale using the multi-beam CEA-CNRS facility JANNUS. JANNUS allows

  4. Alloys having improved resistance to hydrogen embrittlement

    Kane, R.D.; Greer, J.B.; Jacobs, D.F.; Berkowitz, B.J.

    1983-01-01

    The invention involves a process of improving the hydrogen embrittlement resistance of a cold-worked high yield strength nickel/cobalt base alloy containing chromium, and molybdenum and/or tungsten and having individual elemental impurity concentrations as measured by Auger spectroscopy at the crystallographic boundaries of up to about 1 Atomic percent. These elemental impurities are capable of becoming active and mobile at a temperature less than the recrystallization temperature of the alloy. The process involves heat treating the alloy at a temperature above 1300 degrees F but below the temperature of recrystallization for a time of from 1/4 to 100 hours. This is sufficient to effect a reduction in the level of the elemental impurities at the crystallographic boundaries to the range of less than 0.5 Atomic percent without causing an appreciable decrease in yield strength

  5. Ion beam irradiation effects on aromatic polymers

    Shukushima, Satoshi; Ueno, Keiji

    1995-01-01

    We studied the optical and thermal properties of aromatic polymer films which had been irradiated with 1 MeV H + , H 2 + and He + ions. The examined aromatic polymers were polyetherether ketone(PEEK), polyetherimide(PEI), polyether sulfon(PES), polysulfon(PSF), and polyphenylene sulfide(PPS). The optical densities at 300nm of PES and PSF greatly increased after the irradiation. The optical densities at 400nm of all the examined polymer lineally increased with the irradiation dose. The PEEK film which had been irradiated with 1 MeV H + was not deformed above melting point. This demonstrates that cross-linking occurs in PEEK films by ion beam irradiation. As for the effects, depending on the mass of the irradiated ions, it was found that the ions with a high mass induced larger effects on the aromatic polymers for the same absorption energy. (author)

  6. Study of irradiation effect on curcuma polyphenols

    Rejeb, Imen

    2008-01-01

    The present study was carried out to evaluate the effect of gamma irradiation on curcumin (Curcuma Longa rhizome) component, particularly the polyphenolic fraction. Powdered rhizome was irradiated at 0, 5, 10 and 15 KGy (dose rate of 6 KGy / H). Polyphenolics were extracted and total polyphenols conent (TPC) was quantified using the Folin-Ciocalteau method. The irradiation effect was also evaluated by the HPLC technique. The chromatographic analysis showed that the irradiated and non-irradiated curcumin spectrum gave similar data. The antioxidant and antibacterial activities of the phenolic extracts were also assessed. the anti oxidative potential of the sample was evaluated using two radical scavenging methods with DPPH and ABTS. The antimicrobial analysis showed that the phenolic extracts of curcumin inhibited the growth of the studied microorganisms. Our results showed that irradiated samples were not affected in terms of polyphenols content and characteristics. (Author)

  7. Modification of the grain boundary microstructure of the austenitic PCA stainless steel to improve helium embrittlement resistance

    Maziasz, P.J.; Braski, D.N.

    1986-01-01

    Grain boundary MC precipitation was produced by a modified thermal-mechanical pretreatment in 25% cold worked (CW) austenitic prime candidate alloy (PCA) stainless steel prior to HFIR irradiation. Postirradiation tensile results and fracture analysis showed that the modified material (B3) resisted helium embrittlement better than either solution annealed (SA) or 25% CW PCA irradiated at 500 to 600 0 C to approx.21 dpa and 1370 at. ppM He. PCA SA and 25% CW were not embrittled at 300 to 400 0 C. Grain boundary MC survives in PCA-B3 during HFIR irradiation at 500 0 C but dissolves at 600 0 C; it does not form in either SA or 25% CW PCA during similar irradiation. The grain boundary MC appears to play an important role in the helium embrittlement resistance of PCA-B3

  8. Irradiation effect on animal feeds and feedstuffs

    Kume, Tamikazu

    1983-10-01

    Aiming to secure the safety of animal feeds and develop the new resources, the effect of γ-irradiation on disinfection and the changes in components were investigated. Salmonellae and coliforms contaminating in animal feeds and feedstuffs were eliminated by 0.5 -- 0.6 Mrad and 0.5 -- 0.8 Mrad, and osmophilic moulds were sterilized by 0.7 -- 0.75 Mrad. From these results, it is concluded that the dose for disinfection of animal feeds is 0.8 Mrad. The main components were hardly changed by irradiation up to 5 Mrad, and the component changes in irradiated samples could be suppressed during storage while the components in unirradiated samples were markedly changed with the growth of osmophilic moulds. Histamine and lysinoalanine, which may cause the feed poisoning, were never accumulated in feedstuffs by irradiation. The nutritional value of chick feeds was not changed by 1.0 Mrad irradiation. From these results, it is considered that no problem for wholesomeness of animal feeds occurs by irradiation. Therefore, the irradiation is effective for disinfection and keeping the nutritional value of animal feeds during storage. Irradiation promotes the recovery of proteins in the wastewater by coagulation of proteins and improves the property of coagulants due to the degradation of polysaccharides. These results indicate that irradiation is effective to develop the new resources for animal feeds. (author)

  9. Effect of irradiation on the Porphyromonas gingivalis

    Lee, Chang Hee; Kim, Gyu Tae; Choi, Yong Suk; Hwang, Eui Hwan

    2008-01-01

    The aim of this study was to observe a direct effect of irradiation on the periodontopathic Porphyromonas gingivalis (P. gingivalis). P. gingivalis 2561 was exposed to irradiation with a single absorbed dose of 10, 20, 30, and 40 Gy. Changes in viability and antibiotic sensitivity, morphology, transcription, and protein profile of the bacterium after irradiation were examined by pour plating method, disc diffusion method, transmission electron microscopy, RT-PCR, and immunoblot, respectively. Viability of irradiated P. gingivalis drastically reduced as irradiation dose was increased. Irradiated P. gingivalis was found to have become more sensitive to antibiotics as radiation dose was increased. With observation under the transmission electron microscope, the number of morphologically abnormal cells was increased with increasing of irradiation dose. In RT-PCR, decrease in the expression of fim A and sod was observed in irradiated P. gingivalis. In immunoblot, change of profile in irradiated P. gingivalis was found in a number of proteins including 43-kDa fimbrillin. These results suggest that irradiation may affect the cell integrity of P. gingivalis, which is manifested by the change in cell morphology and antibiotic sensitivity, affecting viability of the bacterium.

  10. Effect of irradiation on the Porphyromonas gingivalis

    Lee, Chang Hee; Kim, Gyu Tae; Choi, Yong Suk; Hwang, Eui Hwan [School of Dentistry, Kyung Hee University, Seoul (Korea, Republic of)

    2008-03-15

    The aim of this study was to observe a direct effect of irradiation on the periodontopathic Porphyromonas gingivalis (P. gingivalis). P. gingivalis 2561 was exposed to irradiation with a single absorbed dose of 10, 20, 30, and 40 Gy. Changes in viability and antibiotic sensitivity, morphology, transcription, and protein profile of the bacterium after irradiation were examined by pour plating method, disc diffusion method, transmission electron microscopy, RT-PCR, and immunoblot, respectively. Viability of irradiated P. gingivalis drastically reduced as irradiation dose was increased. Irradiated P. gingivalis was found to have become more sensitive to antibiotics as radiation dose was increased. With observation under the transmission electron microscope, the number of morphologically abnormal cells was increased with increasing of irradiation dose. In RT-PCR, decrease in the expression of fim A and sod was observed in irradiated P. gingivalis. In immunoblot, change of profile in irradiated P. gingivalis was found in a number of proteins including 43-kDa fimbrillin. These results suggest that irradiation may affect the cell integrity of P. gingivalis, which is manifested by the change in cell morphology and antibiotic sensitivity, affecting viability of the bacterium.

  11. Effect of irradiation on the streptococcus mutans

    Ahn, Ki Dong; Kim, Gyu Tae; Choi, Yong Suk; Hwang, Eui Hwan

    2007-01-01

    To observe direct effect of irradiation on cariogenic Streptococcus mutans. S. mutans GS5 was exposed to irradiation with a single absorbed dose of 10, 20, 30, and 40 Gy. Viability and changes in antibiotic sensitivity, morphology, transcription of virulence factors, and protein profile of bacterium after irradiation were examined by pour plate, disc diffusion method, Transmission electron microscopy. RT-PCR, and SDS-PAGE, respectively. After irradiation with 10 and 20 Gy, viability of S. mutans was reduced. Further increase in irradiation dose, however, did not affect the viability of the remaining cells of S. mutans. Irradiated S. mutans was found to have become sensitive to antibiotics. In particular, the bacterium irradiated with 40 Gy increased its susceptibility to cefotaxime, penicillin, and tetracycline. Under the transmission electron microscope, number of morphologically abnormal cells was increased as the irradiation dose was increased. S. mutans irradiated with 10 Gy revealed a change in the cell wall and cell membrane. As irradiation dose was increased. a higher number of cells showed thickened cell wall and cell membrane and lysis, and appearance of ghost cells was noticeable. In RT-PCR, no difference was detected in expression of gtfB and spaP between cells with and without irradiation of 40 Gy. In SDS-PAGE, proteins with higher molecular masses were gradually diminished as irradiation dose was increased. These results suggest that irradiation affects the cell integrity of S. mutans, as observed by SDS-PAGE, and as manifested by the change in cell morphology, antibiotic sensitivity, and eventually viability of the bacterium

  12. Effect of irradiation on the streptococcus mutans

    Ahn, Ki Dong; Kim, Gyu Tae; Choi, Yong Suk; Hwang, Eui Hwan [Kyung Hee Univ., Seoul (Korea, Republic of)

    2007-03-15

    To observe direct effect of irradiation on cariogenic Streptococcus mutans. S. mutans GS5 was exposed to irradiation with a single absorbed dose of 10, 20, 30, and 40 Gy. Viability and changes in antibiotic sensitivity, morphology, transcription of virulence factors, and protein profile of bacterium after irradiation were examined by pour plate, disc diffusion method, Transmission electron microscopy. RT-PCR, and SDS-PAGE, respectively. After irradiation with 10 and 20 Gy, viability of S. mutans was reduced. Further increase in irradiation dose, however, did not affect the viability of the remaining cells of S. mutans. Irradiated S. mutans was found to have become sensitive to antibiotics. In particular, the bacterium irradiated with 40 Gy increased its susceptibility to cefotaxime, penicillin, and tetracycline. Under the transmission electron microscope, number of morphologically abnormal cells was increased as the irradiation dose was increased. S. mutans irradiated with 10 Gy revealed a change in the cell wall and cell membrane. As irradiation dose was increased. a higher number of cells showed thickened cell wall and cell membrane and lysis, and appearance of ghost cells was noticeable. In RT-PCR, no difference was detected in expression of gtfB and spaP between cells with and without irradiation of 40 Gy. In SDS-PAGE, proteins with higher molecular masses were gradually diminished as irradiation dose was increased. These results suggest that irradiation affects the cell integrity of S. mutans, as observed by SDS-PAGE, and as manifested by the change in cell morphology, antibiotic sensitivity, and eventually viability of the bacterium.

  13. Effect of helium irradiation on fracture modes

    Hanamura, T.; Jesser, W.A.

    1982-01-01

    The objective of this work is to determine the crack opening mode during in-situ HVEM tensile testing and how it is influenced by test temperature and helium irradiation. Most cracks were mixed mode I and II. However, between 250 0 C and room temperature the effect of helium irradiation is to increase the amount of mode I crack propagation. Mode II crack opening was observed as grain boundary sliding initiated by a predominantly mode I crack steeply intersecting the grain boundary. Mode II crack opening was absent in irradiated specimens tested between 250 0 C and room temperature, but could be restored by a post irradiation anneal

  14. Irradiation Effects Test Series: Test IE-3. Test results report. [PWR

    Farrar, L. C.; Allison, C. M.; Croucher, D. W.; Ploger, S. A.

    1977-10-01

    The objectives of the test reported were to: (a) determine the behavior of irradiated fuel rods subjected to a rapid power increase during which the possibility of a pellet-cladding mechanical interaction failure is enhanced and (b) determine the behavior of these fuel rods during film boiling following this rapid power increase. Test IE-3 used four 0.97-m long pressurized water reactor type fuel rods fabricated from previously irradiated fuel. The fuel rods were subjected to a preconditioning period, followed by a power ramp to 69 kW/m at a coolant mass flux of 4920 kg/s-m/sup 2/. After a flow reduction to 2120 kg/s-m/sup 2/, film boiling occurred on the fuel rods. One rod failed approximately 45 seconds after the reactor was shut down as a result of cladding embrittlement due to extensive cladding oxidation. Data are presented on the behavior of these irradiated fuel rods during steady-state operation, the power ramp, and film boiling operation. The effects of a power ramp and power ramp rates on pellet-cladding interaction are discussed. Test data are compared with FRAP-T3 computer model calculations and data from a previous Irradiation Effects test in which four irradiated fuel rods of a similar design were tested. Test IE-3 results indicate that the irradiated state of the fuel rods did not significantly affect fuel rod behavior during normal, abnormal (power ramp of 20 kW/m per minute), and accident (film boiling) conditions.

  15. Effect of irradiation on vitamins

    Kilcast, D.

    1994-01-01

    Food irradiation is a physical process involving treatment of food with ionising radiation. Its main uses are reduction in spoilage and pathogenic organisms, inhibition of ripening and sprouting processes, and insect disinfestation. Chemical changes in the treated foods are small, and expert committees have concluded that they carry no special nutritional problems. Some vitamins are sensitive to irradiative degradation, however, and opponents of the process have claimed that extensive destruction will occur. Irradiation doses will, however, be limited by organoleptic changes, and maximum levels are being introduced into legislation for specific foods. Examination of the published literature shows that vitamins C and B 1 are the most sensitive water-soluble vitamins, and that E and A are the most sensitive fat-soluble vitamins. Vitamin losses on irradiation of permitted foods in western countries will not be of nutritional importance. (Author)

  16. Effect of irradiation damage and helium on the swelling and structure of vanadium-base alloys

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1993-12-01

    Swelling behavior and microstructural evolution of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were investigated after irradiation at 420--600C up to 114 dpa. The alloys exhibited swelling maxima between 30 and 80 dpa and swelling decreased on irradiation to higher dpa. This is in contrast to the monotonically increasing swelling of binary alloys that contain Fe, Ni, Cr, Mo, W, and Si. Precipitation of dense Ti 5 Si 3 promotes good resistance to swelling of the Ti-containing alloys and it was concluded that Ti of >3 wt.% and 400--1000 wppm Si are necessary to effectively suppress swelling. Swelling was minimal in V-4Cr-4Ti, identified as the most promising alloy based on good mechanical properties and superior resistance to irradiation embrittlement. V-20Ti doped with B exhibited somewhat higher swelling because of He generation. Lithium atoms, generated from transmutation of 10 B, formed γ-LiV 2 O 5 precipitates and did not seem to produce undesirable effects on mechanical properties

  17. Food irradiation and its biological effects

    Shah, Alok; Nanjappa, C.; Chauhan, O.P.

    2014-01-01

    Irradiation of foods drew attention mostly in 1960s for disinfestation of food grains, spices and sprout inhibition in mainly potato and onion. γ-irradiation at 0.25 to 1 kGy dosage levels are usually used for irradiating grains, legumes, spices and sprout-prone vegetables. Irradiation of foods with in permissible dosage levels of 0.25 to 5 kGy is usually considered fairly safe from human consumption point of view not withstanding usual health concerns about its usage in foods. Irradiation of foods, in mostly solid or semi-solid form, at 5 kGy levels of γ-irradiation can achieve radicidation or, radiation equivalent of pasteurization and, if γ-irradiation is used at 10 kGy, it can achieve radappertization or, radiation equivalent of thermal commercial sterilization. However, the food industry uses γ-irradiation at 0.25 to 2 kGy only for mostly disinfestation of food grains/legumes, spices, sprout inhibition in potato and onion and, for surface sanitation of frozen fish, poultry and meat. Exposure to irradiation creates free radicals in foods that are capable of destroying some of the spoilage and pathogenic microflora but the same can also damage vitamins and enzymes besides creating some new harmful new chemical species, called unique radiolytic products (URPs), by combining with certain chemicals that a food may be laced with (like pesticides/fungicides). Exposure to high-energy electron beams are also known to create deleterious biological effects which may even lead to detection of trace amounts of radioactivity in the food. Some possible causes delineated for such harmful biological effects of irradiation include: irradiation induced vitamin deficiencies, the inactivity of enzymes in the foods, DNA damage and toxic radiolytic products in the foods. Irradiation, a non-thermal food preservation technique, has a role in salvaging enormous post harvest losses (25-30%) in developing economies to increase the per capita availability of foods. (author)

  18. Grain size effect on the mechanical properties of neutron irradiated niobium

    Gusev, M. N.; Maksimkin, O.P.

    2000-01-01

    Samples for mechanical tests were prepared from niobium of technical purity and have form of plates (10·3.5 ·0.3mm) with grain size from 2 to 100 mcm. Neutron irradiation was carried out at the reactor WWR-K to the fluence of 2·10 22 n/m 2 ( Angstroem >0.1 MeV). Tests on uniaxial tension at 293K were performed at the facility, evolving Calvet's microcalorimeter and miniature rapture machine. The developed technique enabled to record heat effects just during the deformation process. As experimental results the characteristics of strength and ductility were defined, as well as values of the latent energy E s , accumulated in material in the process of its deformation up to the moment of destruction. It was found that irradiation of niobium with large-grain structure by neutrons leads to increasing of strength characteristics (yield strength σ 0 .2 changes from 130 to 210 MPa, time-resistance σ b from 200 to 230 MPa) and decreasing of ductility from 36 to 28%. As this takes place the capability of the material to accumulate and dissipate energy of plastic deformation suffers substantial change. There were revealed some additional effects, for instance, the radiation annealing hardening (RAH) (i.e. additional change of properties of irradiated material at annealing), whose maximum takes place at 473K. Its temperature and kinetic parameters were determined in this work. Decreasing of grain size usually leads to decreasing of strengthening under irradiation and to decreasing of RAH effect intensity at subsequent annealing. At the same time decreasing of radiation embrittlement is observed. Consequently, creation of fine-grain structure for some cases can favored the stability of material's properties under irradiation. The obtained results are discussed in context of views on grain boundaries as a defect sink. The relation 'grain boundary volume - grain matrix volume', its influence on RAH-effect and value of latent energy are considered

  19. Effects of mediastinal irradiation on oesophageal function

    Yeoh, E.; Holloway, R.H.; Russo, A.; Tippett, M.; Bermingham, H.; Chatterton, B.; Horowitz, M. [Royal Adelaide Hospital, SA (Australia)

    1996-02-01

    Although it is well recognised that oesophageal symptoms are common during therapeutic irradiation of intrathoracic malignant diseases, the effects of mediastinal irradiation on oesophageal function are poorly defined. To clarify the pathogenesis of these sequelae a prospective study was performed to document comprehensively the effects of mediastinal irradiation on oesophageal function. Oesophageal symptoms, barium swallow, endoscopy, and combined radionuclide scintigraphy and oesophageal manometry were evaluated in eight patients with potentially curable intrathoracic malignant disease before treatment, during the last week of mediastinal irradiation, and six to eight weeks after its completion. Before irradiation, structural abnormalities were excluded by barium swallow and endoscopy. All but one patient experienced odynophagia or dysphagia, or both, during mediastinal irradiation (p<0.001) but endoscopic abnormalities were observed in only three patients and there was no correlation between oesophageal symptoms and endoscopic changes. Irradiation, however, had no significant effect on oesophageal motility or transit. It is concluded that oesophageal symptoms which develop during mediastinal irradiation are not a result of altered oesophageal motility or transit and may reflect increased mucosal sensitivity. (author).

  20. Current limitations of trend curve analysis for the prediction of reactor PV embrittlement

    Gold, R.; McElroy, W.N.

    1986-02-01

    In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integrity is a significant economic consideration since the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant

  1. Effects of gamma irradiation on wheat quality

    Ozkaya, B.; Koksel, H.; Ozkaya, H.; Tutluer, H.

    1994-01-01

    Effect of gamma irradiation at the doses of 2.5,5.0,7.5,10.0 and 12.5 kGy on two bread wheat samples (Bezostaya and Gerek) with distinct physical and technological properties was investigated in this study.Irradiation at the levels used had no significant effect on the flour yields of both varieties.No apparent changes were observed in ash,protein and wet gluten contents of the irradiated samples and control.However,as the radiation level was increased the falling number and sedimentation values of the irradiated samples showed a steady decrease.Thiamine and riboflavin contents also decreased significantly with irradiation.Farinograph absorption increased with increasing radiation exposure.However, dough development time,stability and valorimeter values decreased as radiation levels increased.Maximum resistance to extension(Rm), resistance at constant deformation (R 5) and area(A) values of extensograms decreased in both varieties as radiation levels increased

  2. Irradiation environment and materials behavior

    Ishino, Shiori

    1992-01-01

    Irradiation environment is unique for materials used in a nuclear energy system. Material itself as well as irradiation and environmental conditions determine the material behaviour. In this review, general directions of research and development of materials in an irradiation environment together with the role of materials science are discussed first, and then recent materials problems are described for energy systems which are already existing (LWR), under development (FBR) and to be realized in the future (CTR). Topics selected are (1) irradiation embrittlement of pressure vessel steels for LWRs, (2) high fluence performance of cladding and wrapper materials for fuel subassemblies of FBRs and (3) high fluence irradiation effects in the first wall and blanket structural materials of a fusion reactor. Several common topics in those materials issues are selected and discussed. Suggestions are made on some elements of radiation effects which might be purposely utilized in the process of preparing innovative materials. (J.P.N.) 69 refs

  3. The low-temperature aging embrittlement in a 2205 duplex stainless steel

    Weng, K.L.; Chen, H.R.; Yang, J.R.

    2004-01-01

    The effect of isothermal treatment (at temperatures ranging between 400 and 500 deg. C) on the embrittlement of a 2205 duplex stainless steel (with 45 ferrite-55 austenite, vol.%) has been investigated. The impact toughness and hardness of the aged specimens were measured, while the corresponding fractography was studied. The results show that the steel is susceptible to severe embrittlement when exposed at 475 deg. C; this aging embrittlement is analogous with that of the ferritic stainless steels, which is ascribed to the degenerated ferrite phase. High-resolution transmission electron microscopy reveals that an isotropic spinodal decomposition occurred during aging at 475 deg. C in the steel studied; the original δ-ferrite decomposed into a nanometer-scaled modulated structure with a complex interconnected network, which contained an iron-rich BCC phase (α) and a chromium-enriched BCC phase (α'). It is suggested that the locking of dislocations in the modulated structure leads to the severe embrittlement

  4. Radiation embrittlement of WWER 440 pressure vessel steel and of some improved steels by western producers

    Koutsky, J.; Vacek, M.; Stoces, B.; Pav, T.; Otruba, J.; Novosad, P.; Brumovsky, M.

    1982-01-01

    The resistance was studied of Cr-Mo-V type steel 15Kh2MFA to radiation embrittlement at an irradiation temperature of around 288 degC. Studied was the steel used for the manufacture of the pressure vessel of the Paks nuclear reactor in Hungary. The obtained results of radiation embrittlement and hardening of steel 15Kh2MFA were compared with similar values of Mn-Ni-Mo type steels A 533-B and A 508 manufactured by leading western manufacturers within the international research programme coordinated by the IAEA. It was found that the resistance of steel 15Kh2MFA to radiation embrittlement is comparable with steels A 533-B and A 508 by western manufacturers. (author)

  5. Approach for estimating post-annual reirradiation embrittlement of reactor vessel steels

    Server, W.L.; Taboada, A.

    1985-01-01

    Thermal annealing of a commercial nuclear reactor pressure vessel is a possible solution for extending lifetime in situations where excessive radiation embrittlement has taken place or when the original design life is approached. Two difficult facets of thermal annealing are the degree of toughness recovery after annealing and the post-anneal reirradiation embrittlement behavior. These aspects of annealing are evaluated in this paper by using simple models and translation of the initial irradiation damage curve either vertically or laterally at the point of residual shift after annealing. Results using this methodology are compared to limited actual weld metal measurements of annealing behavior. A forthcoming ASTM Guide on in-place annealing uses this methodology to assess annealing recovery and re-embrittlement response

  6. Radioresistance and immunization effectiveness under internal irradiation

    Kal'nitskij, S.A.

    1978-01-01

    The effect of preliminary immunization on the radioresistance of mice to internal irradiation from incorporated 137 Cs or 90 Sr was studied, and it was found that a preliminary single immunization with bacterial vaccines had a favorable effect on the outcome of radiation injury. The present results suggested that vaccination had a very pronounced radioprotective effect and so may be used as a means of biologic protection from internal irradiation

  7. Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y.

    2014-01-01

    Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10 22 to 3*10 24 n/m 2 depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties), most of the

  8. Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10{sup 22} to 3*10{sup 24} n/m{sup 2} depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties

  9. Proton irradiation effects on organic polymers

    Seguchi, T.; Sasuga, T.; Kawakami, W.; Hagiwara, M.; Kohno, I.; Kamitsubo, H.

    1987-01-01

    Organic polymer films(100 μm thickness) of polyethylene, polypropylene, polyethyleneterephtalate, and polyethersulfone were irradiated by protons of 8 MeV using a cyclotron, and their radiation effects were investigated by the changes of mechanical properties. In order to irradiate protons uniformly over wide area of polymer films, specimens were scanned during proton irradiation using a special apparatus. The absorbed dose was measured by CTA and RCD film dosimeters, and can be determined that 1 μC/cm 2 of 8 MeV proton fluence is equivalent to 54 kGy. For polyethylene and polypropylene, there was no significant difference between proton and electron irradiation for same doses. However, for polyethersulfone the decay of mechanical property was observed to be less than that of irradiation by electron. (author)

  10. Irradiation effects on organic insulators

    Kasen, M.B.

    1986-01-01

    The overall objective of this work is to contribute to development of organic insulators having the cryogenic neutron irradiation resistance required for MFE systems utilizing superconducting magnet confinement. The system for producing standard 3.2-mm (0.125-in) diameter rod specimens discussed in previous reports has been further refined to permit the fabrication of both fiber-reinforced and heat-resin specimens from hot-melt resin systems. The method has been successfully used to produce very high quality specimens duplicating the G-11CR system and specimens from a variant of that system eliminating a boron-containing additive. We have also produced specimens from an epoxy system suitable for impregnation or potting operations and from a bismaleimide polyimide system. These materials will be used in the first irradiation program in the National Low Temperature Neutron Irradiation Facility (NLTNIF) reactor at Oak Ridge. We have refined the 4-K torsional shear test method for evaluating radiation degradation of the fiber-matrix interface and have developed a method of quantitatively measuring changes in fracture energy as a function of radiation dose. Cooperative work with laboratories in Japan and England in this area is continuing and plans are being formulated for joint production, irradiation, and testing of specimens

  11. Radiation hardening and embrittlement of some refractory metals and alloys

    Fabritsiev, S.; Pokrovskyb

    2007-01-01

    Tungsten is proposed for application in the ITER divertor and limiter as plasma facing material. The tungsten operation temperature in the ITER divertor is relatively high. Hence, the ductile properties of tungsten will be controlled by the low temperature radiation embrittlement. The mechanism of radiation hardening and embrittlement under neutron irradiation at low temperature is well studied for FCC metals, in particular for copper. At the same time, low-temperature radiation hardening of BCC materials, in particular for refractory metals, is less studied. This study presents the results of investigation into radiation hardening and embrittlement of pure metals: W, Mo and Nb, and W-Re and Ta-4W alloys. The materials were in the annealed conditions. The specimens were irradiated in the SM-2 reactor to doses of 10 -4 -10 -1 dpa at 80 C and then tested for tension at 80 C. The study of the stress-strain curves of unirradiated specimens revealed a yield drop for W, Mo, Nb, Ta-4W, W-Re. After the yield drop some metals (Mo,Nb) retain their capability for strain hardening and demonstrate a high elongation (20-50%). Radiation hardening is maximum in Mo (∝400MPa) and minimum in Nb (∝100 MPa). In this case the dependence slope for Nb is similar to that for pure copper irradiated in SM-2 under the same conditions. Ii and Ta-4W have a higher slope. Measurement of electrical resistivity of irradiated specimens showed that for all materials it is increased monotonously with an increase in the irradiation dose. A minimum gain in electrical resistivity with a dose was observed for Nb (∝3% at 0.1 dpa). As for Mo it was essentially higher, i.e. ∝ 30%. The gain was maximum for W-Re alloy. Comparison of radiation hardening dose dependencies obtained in this study with the data for FCC metals (Cu) showed that in spite of the quantitative difference the qualitative behavior of these two classes of metals is similar. (orig.)

  12. Hydrogen embrittlement due to hydrogen-inclusion interactions

    Yu, H.Y.; Li, J.C.M.

    1976-01-01

    Plastic flow around inclusions creates elastic misfit which attracts hydrogen towards the regions of positive dilatation. Upon decohesion of the inclusion-matrix interface, the excess hydrogen escapes into the void and can produce sufficient pressure to cause void growth by plastic deformation. This mechanism of hydrogen embrittlement can be used to understand the increase of ductility with temperature, the decrease of ductility with hydrogen content, and the increase of ductility with the ultimate strength of the matrix. An examination of the effect of the shape of spheroid inclusion reveals that rods are more susceptible to hydrogen embrittlement than disks. The size of the inclusion is unimportant while the volume fraction of inclusions plays the usual role

  13. The strengthening of embrittled books using gamma radiation

    Egan, A.; Mardian, J.; Foot, M.; King, E.; Millington, A.; Nevin, M.; Butler, C.; Barker, J.; Fletcher, D.

    1995-01-01

    The embrittlement of papers, manufactured through processes introduced in the mid-19th century, has caused many millions of books to become fragile, even to the point of being unusable. During the 1980s the British Library funded a research programme, carried out at the University of Surrey, to develop a technology which could be used to treat brittle books on a large scale, with the goal of greatly extending their useful life. The process developed, known as graft co-polymerization, involves three stages: i) application of a cocktail of monomers to the book's pages; ii) equilibration of these monomers throughout the text block; and iii) a low, slow dose of γ-radiation to effect polymerization. In collaboration with the British Library, Nordion International has designed a full-scale book-strengthening plant capable of processing between 200,000 and 500,000 and 500,000 books per year, with estimated prices to customers in the region of 1 8-10 per volume (US $12-16). In order to test the equipment and procedures that would be involved in such a plant, pilot-scale equipment has been designed and assembled on the premises of Isotron plc, where use is made of a conventional irradiator. This paper gives details of the graft co-polymerization process, and some results of the pilot-scale work, in terms of both efficacy and controllability. It also discusses the technical and economic feasibility of building and running a full-scale plant. (author)

  14. Study of intergranular embrittlement in Fe-12Mn alloys

    Lee, H.J.

    1982-06-01

    A high resolution scanning Auger microscopic study has been performed on the intergranular fracture surfaces of Fe-12Mn steels in the as-austenitized condition. Fracture mode below the ductile-brittle transition temperature was intergranular whenever the alloy was quenched from the austenite field. The intergranular fracture surface failed to reveal any consistent segregation of P, S, As, O, or N. The occasional appearance of S or O on the fracture surface was found to be due to a low density precipitation of MnS and MnO 2 along the prior austenite boundaries. An AES study with Ar + ion-sputtering showed no evidence of manganese enrichment along the prior austenite boundaries, but a slight segregation of carbon which does not appear to be implicated in the tendency toward intergranular fracture. Addition of 0.002% B with a 1000 0 C/1h/WQ treatment yielded a high Charpy impact energy at liquid nitrogen temperature, preventing the intergranular fracture. High resolution AES studies showed that 3 at. % B on the prior austenite grain boundaries is most effective in increasing the grain boundary cohesive strength in an Fe-12Mn alloy. Trace additions of Mg, Zr, or V had negligible effects on the intergranular embrittlement. A 450 0 C temper of the boron-modified alloys was found to cause tempered martensite embrittlement, leading to intergranular fracture. The embrittling treatment of the Fe-12Mn alloys with and without boron additions raised the ductile-brittle transition by 150 0 C. This tempered martensite embrittlement was found to be due to the Mn enrichment of the fracture surface to 32 at. % Mn in the boron-modified alloy and 38 at. % Mn in the unmodified alloy. The Mn-enriched region along the prior austenite grain boundaries upon further tempering is believed to cause nucleation of austenite and to change the chemistry of the intergranular fracture surfaces. 61 figures

  15. Effect of irradiation on foodstuffs Pt. 4

    Kluender, U.; Boegl, W.

    1980-01-01

    In the present study of the relevant literature the results of irradiation experiments with 32 foodstuffs have been compiled and discussed. This study is intented to give a survey on chemical changes in irradiated food, and neither microbiological nor toxicological and physiological aspects were taken into account. The results published by the authors of the original papers have been compiled in form of a dictionary which contains all important data such as radiation source, irradiation conditions, treatment and characteristics of the sample, investigation methods, results of the chemical and organoleptical changes etc. In addition, the effects of irradiation both on individual food substances and individual groups of foodstuffs have been summarized. Furthermore, the effects of irradiation on sensory characteristics and the atmospheric influence during irradiation are given seperately. The last chapter contains a comparison between the chemical changes of food due to irradiation treatment and those caused by conventional methods. The final discussion of the results will be published seperately. (orig./MG) [de

  16. Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

    Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-05-15

    Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, {sigma}{sub y}={sigma}{sub i}+k{sub y} d{sup -1/2} between lower yield stress, {sigma}{sub y}, and grain size, 2d, the information about the effect of irradiation on the parameters {sigma}{sub i} and k{sub y} is obtained. Taking as a base interpretation of {sigma}{sub i} and k{sub y} given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of

  17. Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

    Milasin, N.

    1964-05-01

    Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, σ y =σ i +k y d -1/2 between lower yield stress, σ y , and grain size, 2d, the information about the effect of irradiation on the parameters σ i and k y is obtained. Taking as a base interpretation of σ i and k y given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of the experimental results obtained the relative microstructure and

  18. Combined effect of external and internal irradiation

    Kiradzhiev, G.

    1987-01-01

    Some of the general regularities of the combined effect of external irradiation and iodine 131 are discussed. Data are adduced showing that modification of the effects of these two radiation factors, when jointly applied, is also determined by the quantitative relations of the applied doses of external and internal irradiation, referred to a particular moment of the effects. It was shown that the effects of the radionuclides in these combined radiation injuries are basically realized by two mechanisms: 1. changes are found in the radionuclide kinetic parameters (nonspecific effects); 2. changes in their kinetic parameters are absent (specific effect). These two mechanisms underlie different approaches to therapy

  19. Irradiation, annealing, and reirradiation research in the ORNL heavy-section steel irradiation program

    Nanstad, R.K.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.

    1997-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results from work performed as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. The HSSI Program focuses on annealing and re-embrittlement response of materials which are representative of those in commercial RPVs and which are considered to be radiation-sensitive. Experimental studies include (1) the annealing of materials in the existing inventory of previously irradiated materials, (2) reirradiation of previously irradiated/annealed materials in a collaborative program with the University of California, Santa Barbara (UCSB), (3) irradiation/annealing/reirradiation of U.S. and Russian materials in a cooperative program with the Russian Research Center-Kurchatov Institute (RRC-KI), (4) the design and fabrication of an irradiation/anneal/reirradiation capsule and facility for operation at the University of Michigan Ford Reactor, (5) the investigation of potential for irradiation-and/or thermal-induced temper embrittlement in heat-affected zones (HAZs) of RPV steels due to phosphorous segregation at grain boundaries, and (6) investigation of the relationship between Charpy impact toughness and fracture toughness under all conditions of irradiation, annealing, and reirradiation

  20. Present status of the disk pressure tests for hydrogen embrittlement

    Fidelle, J.P.

    1985-05-01

    The Disk Pressure Tests (DPT) have been developed considerably theoretically and experimentally for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for Environment Embrittlement due to H 2 , hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably for pressure up to 300 MPa and temperature (-160 0 C to 1000 0 C). Very low strain rate -longer than a month- tests have been able to evidence embrittlement of FFC alloys where H diffusivity is low. Conversely for very oxidation - sensitive metals (e.g. Nb and Ta) effects may appear only at somewhat high rates. The relationship between dynamic (increasing stress) tests, static (delayed failure) and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analyzed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 150 + materials in different conditions. From the tests on a large number of metal systems, a theory of HE has been derived which accounts for the behavior of metals and alloys either embrittled and or hydrited. Finally comparison of HGE tests and service behavior of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service

  1. Hydride embrittlement in zircaloy components

    Lobo, Raquel M.; Andrade, Arnaldo H.P.; Castagnet, Mariano, E-mail: rmlobo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Zirconium alloys are used in nuclear reactor cores under high-temperature water environment. During service, hydrogen is generated by corrosion processes, and it is readily absorbed by these materials. When hydrogen concentration exceeds the terminal solid solubility, the excess hydrogen precipitates as zirconium hydride (ZrH{sub 2}) platelets or needles. Zirconium alloys components can fail by hydride cracking if they contain large flaws and are highly stressed. Zirconium alloys are susceptible to a mechanism for crack initiation and propagation termed delayed hydride cracking (DHC). The presence of brittle hydrides, with a K{sub Ic} fracture toughness of only a few MPa{radical}m, results in a severe loss in ductility and toughness when platelet normal is oriented parallel to the applied stress. In plate or tubing, hydrides tend to form perpendicular to the thickness direction due to the texture developed during fabrication. Hydrides in this orientation do not generally cause structural problems because applied stresses in the through-thickness direction are very low. However, the high mobility of hydrogen in a zirconium lattice enables redistribution of hydrides normal to the applied stress direction, which can result in localized embrittlement. When a platelet reaches a critical length it ruptures. If the tensile stress is sufficiently great, crack initiation starts at some of these hydrides. Crack propagation occurs by repeating the same process at the crack tip. Delayed hydride cracking can degrade the structural integrity of zirconium alloys during reactor service. The paper focuses on the fracture mechanics and fractographic aspects of hydride material. (author)

  2. Effects of ion beam irradiation on semiconductor devices

    Nashiyama, Isamu; Hirao, Toshio; Itoh, Hisayoshi; Ohshima, Takeshi [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    1997-03-01

    Energetic heavy-ion irradiation apparatus has been developed for single-event effects (SEE) testing. We have applied three irradiation methods such as a scattered-ion irradiation method, a recoiled-atom irradiation method, and a direct-beam irradiation method to perform SEE testing efficiently. (author)

  3. Empirical correlation between mechanical and physical parameters of irradiated pressure vessel steels

    Tipping, P.; Solt, G.; Waeber, W.

    1991-02-01

    Neutron irradiation embrittlement of nuclear reactor pressure vessel (PV) steels is one of the best known ageing factors of nuclear power plants. If the safety limits set by the regulators for the PV steel are not satisfied any more, and other measures are too expensive for the economics of the plant, this embrittlement could lead to the closure of the plant. Despite this, the fundamental mechanisms of neutron embrittlement are not yet fully understood, and usually only empirical mathematical models exist to asses neutron fluence effects on embrittlement, as given by the Charpy test for example. In this report, results of a systematic study of a French forging (1.2 MD 07 B), irradiated to several fluences will be reported. Mechanical property measurements (Charpy tensile and Vickers microhardness), and physical property measurements (small angle neutron scattering - SANS), have been done on specimens having the same irradiation or irradiation-annealing-reirradiation treatment histories. Empirical correlations have been established between the temperature shift and the decrease in the upper shelf energy as measured on Charpy specimens and tensile stresses and hardness increases on the one hand, and the size of the copper-rich precipitates formed by the irradiation on the other hand. The effect of copper (as an impurity element) in enhancing the degradation of mechanical properties has been demonstrated; the SANS measurements have shown that the size and amount of precipitates are important. The correlations represent the first step in an effort to develop a description of neutron irradiation induced embrittlement which is based on physical models. (author) 6 figs., 27 refs

  4. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  5. Progress in identification of radiation embrittlement mechanisms

    Hawthorne, J.R.

    1988-01-01

    This report outlines recent advances in the isolation and understanding of mechanisms behind known composition influences on he radiation embrittlement sensitivity of reactor pressure vessel steels at 288 deg. C. The advances are largely the product of joint investigations by Materials Engineering Associates (MEA) and other laboratories in the U.S. and overseas under cooperative and subcontract arrangements. Specific objectives were: confirmation of the suspect Cu mechanism, identification of the process for the Cu:Ni synergism, and isolation of the P mechanism in radiation sensitivity development. The investigations proceeded with MEA-supplied steels and iron alloys from 4-way split laboratory melts; research tools included Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM), Field Ion Microscopy (FIM), Small Angle Neutron Scattering (SANS), Positron Annihilation (PA) and Auger Electron Spectroscopy (AES). Experimental results show that P and Cu enhance the radiation elevation of yield strength and that the associated mechanisms are a radiation-induced precipitation of P or Cu-rich clusters which impede dislocation motion. With high Cu alloys, a Cu phosphide is formed in preference to P precipitates and the P contribution is greatly reduced. Effects of postirradiation annealing and reirradiation are also reported. (author)

  6. Modifying effect of low dose irradiation

    Kalendo, G.S.

    1989-01-01

    It is shown that irradiation of Hela cells with stimulating doses of 0,1 Gy changes the cells' response to the subsequent radiation effect of greater value: instead of DNA synthesis inhibition stimulation takes place. Modifying effect of preliminary irradiation with 0,1 Gy manifests it self only in case if there is a certain time interval not less than 3 minutes and not more than 10 minutes (3-5 minutes is optimal interval). Data on modifying effect with 0,1 Gy at subcellular and cellular-population levels are presented. 21 refs.; 6 figs

  7. Synergistic effects of irradiation of waste water

    Woodbridge, D.D.; Cooper, P.C.; Vandenburg, A.J.; Musselman, H.D.; Lowe, H.N.; Florida Inst. of Tech., Melbourne; Army Facilities Engineering Support Agency, Fort Belvoir, Va.

    1975-01-01

    Theoretical considerations of the use of high level radiation in the treatment of waste water have failed to consider the effects of the hydrated electron and the potential of possible synergistic effects of combining chlorine, oxygen, and irradiation. An extensive testing program at the University Center for Pollution Research of Florida Institute of Technology over the past four years has shown that irradiation of waste water samples immersed in an aqueous environment provide bacterial kill and reduction in organic pollution far greater than that obtained from theoretical considerations of G values and earlier experiments where the waste samples were not immersed in an aqueous environment. These testing programs have investigated the synergistic effects of combining oxygen and irradiation. Each of these combined treatments resulted in an increased bacterial kill factor. Tests on Staphylococcus aureus bacteria and fecal streptococcus bacteria indicate that the synergistic effects observed for fecal coliform bacteria also apply to the pathogenic bacteria. A statistical analysis of the data obtained shows the interrelationships between the various effects on the bacteria. A definite shielding factor due to the turbidity of the waste water has been shown to exist. Synergistic effects have been shown to significantly offset the shielding effects. Optimization of these synergistic effects can greatly increase the effectiveness of irradiation in the treatment of waste water. (orig.) [de

  8. The role of pressure vessel embrittlement in the long term operation of nuclear power plants

    Ballesteros, A.; Ahlstrand, R.; Bruynooghe, C.; Estorff, U. von; Debarberis, L.

    2012-01-01

    Highlights: ► Relevant open scientific issues for the long term operation of RPVs are discussed (flux effect, late blooming phases, etc.). ► Several European and American research programmes dealing with these open issues are reviewed. ► A method for consolidation and preservation of knowledge in this field is presented. - Abstract: The lack of new build of plants over the last twenty years has resulted in a switch within the industry from design, construction and development of new systems to the strengthening of safety systems and to the life extension, or long term operation (LTO), of existing reactors. The most relevant component of any nuclear power plan (NPP) is the reactor pressure vessel (RPV). This is because currently the RPV is still considered irreplaceable or prohibitively expensive to replace. This means, that if it degrades sufficiently, it could be the operational life limiting feature of the NPP. A RPV operational life of 60 years is being considered frequently by many utilities in their plant life management programmes. Areas of improvement facing long term operation are the reduction of uncertainties in the embrittlement parameters of irradiated vessels, and the development of embrittlement trend curves at high fluence levels, where surveillance data are scarce. Different techniques can be used to upgrade the surveillance programmes, as the use of miniature or reconstituted specimens and the application of best estimate assessment tools (e.g. Master Curve). Several relevant international research projects are on-going or have been proposed to clarify the material condition of long operated vessels. Knowledge management is a complementary tool, but not for it less important. The general context for LTO of RPVs is presented in this paper. Starting with a review of relevant embrittlement issues still open, followed by presenting the different techniques and tools that can be used to support LTO, and summarising the scopes of relevant European

  9. Immediate effect of irradiation on microvasculature

    Krishnan, L.; Krishnan, E.C.; Jewell, W.R.

    1988-01-01

    The immediate effects of irradiation on microvasculature in muscle in an animal model are described in this paper. By using triple isotopes of 125 I, 131 I, and 22 Na, the transcapillary transfer of albumin from the vascular bed to the extravascular space is determined in terms of mg/g of tissue, after single doses of 2 to 14 Gy. These results reveal an increase in the extravascular albumin immediately after irradiation and suggest an instantaneous compromise in vascular permeability even after 2 Gy. This effect was apparently dose related

  10. Influence of a cyclic load on the embrittlement kinetics of alloys by the example of the 475 C embrittlement of duplex steel and the dynamic embrittlement of a nickel base alloy; Einfluss einer zyklischen Belastung auf die Versproedungskinetik von Legierungen am Beispiel der 475 C-Versproedung von Duplexstahl und der dynamischen Versproedung einer Nickelbasislegierung

    Wackermann, Ken

    2015-07-07

    The objective of this study was to investigate the dependence of high temperature embrittlement mechanisms on high temperature fatigue and vice versa. As model embrittlement mechanisms the 475 C Embrittlement of ferritic austenitic duplex stainless steel (1.4462) and the Dynamic Embrittlement of nickel-based superalloys (IN718) were selected. The 475 C Embrittlement is a thermally activated decomposition of the ferritic phase which hardens the material. In contrast to this a cyclic plastic deformation weakens the steel by a deformation-induced dissolution of the decomposition. Fatigue tests with different frequencies, loading amplitudes at room temperature and at 475 C with Duplex Stainless Steel in different states of embrittlement show that the ongoing 475 C Embrittlement and the deformation-induced dissolution are competing mechanisms. It depends on the frequency, the loading amplitude and the temperature which mechanism is dominant. Applying the model of the yield stress distribution function to the hysteresis branches of the fatigue tests allows an analysis of the fatigue behaviour of each phase individually. This analysis shows that the global fatigue behaviour for the test conditions applied in this study is mainly controlled by the ferritic phase. According to the existing understanding of Dynamic Embrittlement it is an oxygen grain boundary diffusion arising by tensile stress at elevated temperatures with the result of a fast intercrystalline crack propagation. In reference tests under vacuum conditions without oxygen grain boundary diffusion, a slow transcrystalline fracture appears. To analyse the Dynamic Embrittlement, the crack propagation was tested at 650 C with different frequencies and superimposed hold times in the fatigue cycle at maximum stress. The results shows that the existing model of Dynamic Embrittlement needs to be adapted to the effects of cyclic plastic deformation. In hold times, the oxygen grain boundary diffusion in front of the

  11. Influence of sulfur, phosphorus, and antimony segregation on the intergranular hydrogen embrittlement of nickel

    Bruemmer, S.M.; Baer, D.R.; Jones, R.H.; Thomas, M.T.

    1983-01-01

    The effectiveness of sulfur, phosphorus, and antimony in promoting the intergranular embrittlement of nickel was investigated using straining electrode tests in 1N H 2 SO 4 at cathodic potentials. Sulfur was found to be the critical grain boundary segregant due to its large enrichment at grain boundaries (10 4 to 10 5 times the bulk content) and the direct relationship between sulfur coverage and hydrogeninduced intergranular failure. Phosphorus was shown to be significantly less effective than sulfur or antimony in inducing the intergranular hydrogen embrittlement of nickel. The addition of phosphoru to nickel reduced the tendency for intergranular fracture and improved ductility because phosphoru segregated strongly to grain interfaces and limited sulfur enrichment. The hydrogen embrittling potency of antimony was also less than that of sulfur while its segregation propensity was considerably less. It was found that the effectiveness of segregated phosphorus and antimony in prompting inter granular embrittlement vs that of sulfur could be expressed in terms of an equivalent grain boundary sulfur coverage. The relative hydrogen embrittling potencies of sulfur, phosphorus, and antimony are discussed in reference to general mechanisms for the effect of impurity segregation on hydrogeninduced intergranular fracture

  12. Proton irradiation effects in silicon devices

    Simoen, E; Vanhellemont, J; Alaerts, A [IMEC, Leuven (Belgium); and others

    1997-03-01

    Proton irradiation effects in silicon devices are studied for components fabricated in various substrates in order to reveal possible hardening effects. The degradation of p-n junction diodes increases in first order proportionally with the fluence, when submitted to 10 MeV proton irradiations in the range 5x10{sup 9} cm{sup -2} to 5x10{sup 11} cm{sup -2}. The damage coefficients for both p- and n-type Czochralski, Float-Zone and epitaxial wafers are reported. Charge-Coupled Devices fabricated in a 1.2 {mu}m CCD-CMOS technology are shown to be quite resistant to 59 MeV H{sup +} irradiations, irrespective of the substrate type. (author)

  13. Recrystallization and embrittlement of sintered tungsten

    Bega, N.D.; Babak, A.V.; Uskov, E.I.

    1982-01-01

    The recrystallization of sintered tungsten with a cellular structure of deformation is studied as related to its embrittlement. It is stated that in case of preliminary recrystallization the sintered tungsten crack resistance does not depend on the testing temperature. The tungsten crack resistance is shown to lower with an increase of the structure tendency to primary recrystallization [ru

  14. Effects of irradiation on platelet function

    Rock, G.; Adams, G.A.; Labow, R.S.

    1988-01-01

    Current medical practice involves the irradiation of blood components, including platelet concentrates, before their administration to patients with severe immunosuppression. The authors studied the effect of irradiation on in vitro platelet function and the leaching of plasticizers from the bag, both immediately and after 5 days of storage. The platelet count, white cell count, pH, glucose, lactate, platelet aggregation and release reaction, and serotonin uptake were not altered by the irradiation of random-donor or apheresis units with 2000 rads carried out at 0 and 24 hours and 5 days after collection. The leaching of di(2-ethylhexyl)phthalate from the plastic bags followed by the conversion to mono(2-ethylhexyl)phthalate was not increased by irradiation. Therefore, it is possible to irradiate platelet concentrates on the day of collection and subsequently store them for at least 5 days while maintaining in vitro function. This procedure could have considerable benefit for blood banks involved in the provision of many platelet products

  15. Evaluation of gamma irradiation effect and Pseudomonas ...

    Antagonistic effect of Pseudomonas fluorescens and influence of gamma irradiation on the development of Penicillium expansum, the causal agent of postharvest disease on apple fruit was studied. P. fluorescens was originally isolated from rhizosphere of the apple trees. Suspension of P. fluorescens and P. expansum ...

  16. Post irradiation effects (PIE) in integrated circuits

    Barnes, C.E.; Shaw, D.C.; Fleetwood, D.M.; Winokur, P.S.

    1992-01-01

    Post Irradiation Effects (PIE) ranging from normal recovery catastrophic failure have been observed in integrated circuits during the PIE period. These variations indicate that a rebound or PIE recipe used for radiation hardness assurance must be chosen with care. In this paper, the authors provide examples of PIE in a variety of integrated circuits of importance to spacecraft electronics

  17. Enhancement effect of irradiation by methotrexate

    Shehata, W.M.; Meyer, R.L.

    1980-01-01

    Three cases are described in which complications developed which were believed to be due to the enhancement effect of irradiation by methotrexate during the course of therapy for lung, kidney, and bladder cancer. These included esophageal and large bowel complications. In two of these cases, the patients improved with conservative therapy

  18. Effect of UV irradiation on cutaneous cicatrices

    Due, Eva; Rossen, Kristian; Sorensen, Lars Tue

    2007-01-01

    The aim of this study was to examine the effect of ultraviolet (UV) irradiation on human cutaneous cicatrices. In this randomized, controlled study, dermal punch biopsy wounds served as a wound healing model. Wounds healed by primary or second intention and were randomized to postoperative solar UV...... postoperatively, UV-irradiated cicatrices healing by second intention: (i) were significantly pointed out as the most disfiguring; (ii) obtained significantly higher scores of colour, infiltration and cicatrix area; and (iii) showed significantly higher increase in skin-reflectance measurements of skin......-pigmentation vs. non-irradiated cicatrices. No histological, immunohistochemical or biochemical differences were found. In conclusion, postoperative UV exposure aggravates the clinical appearance of cicatrices in humans....

  19. Effect of irradiation spectrum on the microstructure of ion-irradiated Al2O3

    Zinkle, S.J.

    1994-01-01

    Polycrystalline samples of alpha-alumina have been irradiated with various ions ranging from 3.6 MeV Fe + to 1 MeV H + ions at 650 C. Cross-section transmission electron microscopy was used to investigate the depth-dependent microstructure of the irradiated specimens. The microstructure following irradiation was observed to be dependent on the irradiation spectrum. In particular, defect cluster nucleation was effectively suppressed in specimens irradiated with light ions such as 1 MeV H + ions. On the other hand, light ion irradiation tended to accelerate the growth rate of dislocation loops. The microstructural observations are discussed in terms of ionization enhanced diffusion processes

  20. The biological effectiveness of antiproton irradiation

    Holzscheiter, Michael H.; Bassler, Niels; Agazaryan, Nzhde

    2006-01-01

    ever measurements of the biological effectiveness of antiprotons. Materials and methods: V79 cells were suspended in a semi-solid matrix and irradiated with 46.7 MeV antiprotons, 48 MeV protons, or 60Co c-rays. Clonogenic survival was determined as a function of depth along the particle beams. Dose...... and particle fluence response relationships were constructed from data in the plateau and Bragg peak regions of the beams and used to assess the biological effectiveness. Results: Due to uncertainties in antiproton dosimetry we defined a new term, called the biologically effective dose ratio (BEDR), which...... has a higher relative biological effectiveness (RBE). Conclusion: We have produced the first measurements of the biological consequences of antiproton irradiation. These data substantiate theoretical predictions of the biological effects of antiproton annihilation within the Bragg peak, and suggest...

  1. Effects of gamma irradiation on deteriorated paper

    Bicchieri, Marina; Monti, Michela; Piantanida, Giovanna; Sodo, Armida

    2016-08-01

    Even though gamma radiation application, also at the minimum dosage required for disinfection, causes depolymerization and degradation of the paper substrate, recently published papers seemed, instead, to suggest that γ-rays application could be envisaged in some conditions for Cultural Heritage original documents and books. In some of the published papers, the possible application of γ-rays was evaluated mainly by using mechanical tests that scarcely reflect the chemical modifications induced in the cellulosic support. In the present article the effect of low dosage γ-irradiation on cellulosic substrates was studied and monitored applying different techniques: colorimetry, spectroscopic measurements, carbonyl content and average viscometric degree of polymerization. Two different papers were investigated, a non-sized, non-filled cotton paper, and a commercial permanent paper. To simulate a real deteriorated document, which could need γ-rays irradiation, some samples were submitted to a hydrolysis treatment. We developed a treatment based on the exposition of paper to hydrochloric acid vapors, avoiding any contact of the samples with water. This method induces a degradation similar to that observed on original documents. The samples were then irradiated with 3 kGy γ-rays at a 5258 Gy/h rate. The aforementioned analyses were performed on the samples just irradiated and after artificial ageing. All tests showed negative effects of gamma irradiation on paper. Non-irradiated paper preserves better its appearance and chemical properties both in the short term and after ageing, while the irradiated samples show appreciable color change and higher oxidation extent. Since the Istituto centrale restauro e conservazione patrimonio archivistico e librario is responsible for the choice of all restoration treatments that could be applied on library and archival materials under the protection of the Italian State (http://www.icpal.beniculturali.it/allegati/DM-7

  2. The influences of impurity content, tensile strength, and grain size on in-service temper embrittlement of CrMoV steels

    Cheruvu, N.S.; Seth, B.B.

    1989-01-01

    The influences of impurity levels, grain size, and tensile strength on in-service temper embrittlement of CrMoV steels have been investigated. The samples for this study were taken from steam turbine CrMoV rotors which had operated for 15 to 26 years. The effects of grain size and tensile strength on embrittlement susceptibility were separated by evaluating the embrittlement behavior of two rotor forgings made from the same ingot after an extended step-cooling treatment. Among the residual elements in the steels, only P produces a significant embrittlement. The variation of P and tensile strength has no effect on in-service temper embrittlement susceptibility, as measured by the shift in fracture appearance transition temperature (FATT). However, the prior austenite grain size plays a major role in service embrittlement. The fine grain steels with a grain size of ASTM No. 9 or higher are virtually immune to in-service embrittlement. In steels having duplex grain sizes, embrittlement susceptibility is controlled by the size of coarser grains. For a given steel chemistry, the coarse grain steel is more susceptible to in-service embrittlement, and a decrease in ASTM grain size number from 4 to 0/1 increases the shift in FATT by 61 degrees C (10/10 degrees F). It is demonstrated that long-term service embrittlement can be simulated, except in very coarse grain steels, by using the extended step-cooling treatment. The results of step-cooling studies show that the coarse grain rotor steels take longer time during service to reach a fully embrittled state than the fine grain rotor steels

  3. Effect of medium and post-irradiation storage on rooting of irradiated onions

    Singh, Rita

    2000-01-01

    Rooting test for detection of irradiation in onion bulbs was studied. Onions were exposed to different dose levels of 30, 60, 90, 120 and 150 Gy. The effects of irradiation dose, cultivar difference, rooting medium and post-irradiation storage on the rooting were investigated. The number and the length of the roots formed in onions were found to decrease on irradiation. The effect was more at higher doses. The effect of irradiation on rooting was also evident after 120 days of storage. (author)

  4. Effect of irradiation on the tensile properties of niobium-base alloys

    Grossbeck, M.L.; Heestand, R.L.; Atkin, S.D.

    1986-11-01

    The alloys Nb-1Zr and PWC-11 (Nb-1Zr-0.1C) were selected as prime candidate alloys for the SP-100 reactor. Since the mechanical properties of niobium alloys irradiated to end-of-life exposure levels of about 2 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures above 1300 K were not available, an irradiation experiment (B-350) in EBR-II was conducted. Irradiation creep, impact properties, bending fatigue, and tensile properties were investigated; however, only tensile properties will be reported in this paper. The tensile properties were studied since they easily reveal the common irradiation phenomena of hardening and embrittlement. Most attention was directed to testing at the irradiation temperature. Further testing was conducted at lower temperatures in order to scope the behavior of the alloys in cooldown conditions

  5. Hydrogen embrittlement of titanium and its alloys - a literature review

    Aho-Mantila, I.; Haemaelaeinen, H.

    1986-05-01

    Hydrogen embrittlement data of titanium and its alloys is reviewed. Especially the results obtained in spent nuclear fuel repository conditions with commercially pure titanium and TiCode-12 alloy are examined. The results show that the mechanical properties of titanium are not much affected by hydrogen when tested by smooth specimens. Much greater effects can be expected with notched fracture mechanics specimens. However, only limeted data is available. Hydrogen distribution in titanium is affected by stress, alloy composition and temperature gradients. In order to model the hydrogen-induced crack growth in titanium much more mechanistic work is needed especially to understand the behaviour of hydrogen in crack tip stress field. (author)

  6. Late effects of irradiation in mouse jejunum

    Reynaud, A.; Travis, E.L.

    1984-01-01

    The response of mouse jejunum at intervals up to 1 year after single 'priming' doses of X-rays has been assessed by crypt survival after retreatment with single doses of X-rays and morphometric analysis of changes in the intestinal submucosa. The crypt dose-survival curves in mice re-irradiated at 2, 6, or 12 months after priming irradiation were displaced to higher doses in pre-treated than in non-pre-treated mice and were characterized by higher D 0 values. Misonidazole given before the test exposure reversed this effect so that the dose survival curve for crypts in pre-treated mice were superimposed on that for mice not previously irradiated, suggesting that the increase in isoeffect dose and the change in the D 0 in previously exposed mice was due to crypt hypoxia. Quantifications of the area of the submucosa showed that its area was increased at all three times after the priming doses and was a result of collagen deposition and oedema. Thus, the hypoxia in the crypts was probably secondary to these changes. Deaths began at 6-7 months after priming irradiation and were due to intestinal obstruction and stenosis. Thus, as in other tissues, two phases of injury can be assayed in the intestine of experimental animals. (author)

  7. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  8. Oxidation-induced embrittlement and structural changes of Zircaloy-4 tubing in steam at 700-1000 deg. C

    Ali, A E; Huessein, A G; El-Sayed, A A; El Banna, O A [Atomic Energy Authority, Cairo (Egypt); El Raghy, S M [Cairo Univ. (Egypt). Faculty of Engineering

    1997-02-01

    The oxidation-induced embrittlement and structural changes of Zircaloy-4 (KWU-Type) tubing was investigated under light water reactors (LWR) Loss-of-Coolant. Accident conditions (LOCA) in temperature range 700-1000 deg. C. The effect of hydrogen addition to steam was also investigated in the temperature range 800-1000 deg. C. The oxidation-induced embrittlement was found to be a function of both temperature and time. Fractography investigation of oxidized tubing showed a typical brittle fracture in the stabilized-alpha zone. The microhardness measurements revealed that the alpha-Zr is harder than that near the mid-wall position. The oxidation-induced embrittlement at 900 deg. C was found to be higher than at 1000 deg. C. The results also indicated that the addition of 5% by volume hydrogen to steam resulted in an increase in the degree of embrittlement. (author). 22 refs, 9 figs, 3 tabs.

  9. Irradiation effects in magnesium and aluminium alloys

    Sturcken, E.F.

    1979-01-01

    Effects of neutron irradiation on microstructure, mechanical properties and swelling of several magnesium and aluminium alloys were studied. The neutron fluences of 2-3 X 10 22 n/cm 2 , >0.2 MeV produced displacement doses of 20 to 45 displacements per atom (dpa). Ductility of the magnesium alloys was severely reduced by irradiation induced recrystallization and precipitation of various forms. Precipitation of transmuted silicon occurred in the aluminium alloys. However, the effect on ductility was much less than for the magnesium alloys. The magnesium and aluminium alloys had excellent resistance to swelling: The best magnesium alloy was Mg/3.0 wt% Al/0.19 wt% Ca; its density decreased by only 0.13%. The best aluminium alloy was 6063, with a density decrease of 0.22%. (Auth.)

  10. Quantification of biologically effective environmental UV irradiance

    Horneck, G.

    To determine the impact of environmental UV radiation on human health and ecosystems demands monitoring systems that weight the spectral irradiance according to the biological responses under consideration. In general, there are three different approaches to quantify a biologically effective solar irradiance: (i) weighted spectroradiometry where the biologically weighted radiometric quantities are derived from spectral data by multiplication with an action spectrum of a relevant photobiological reaction, e.g. erythema, DNA damage, skin cancer, reduced productivity of terrestrial plants and aquatic foodweb; (ii) wavelength integrating chemical-based or physical dosimetric systems with spectral sensitivities similar to a biological response curve; and (iii) biological dosimeters that directly weight the incident UV components of sunlight in relation to the effectiveness of the different wavelengths and to interactions between them. Most biological dosimeters, such as bacteria, bacteriophages, or biomolecules, are based on the UV sensitivity of DNA. If precisely characterized, biological dosimeters are applicable as field and personal dosimeters.

  11. Effect of UV-irradiation on rotavirus

    Smirnov, Y.A.; Kapitulets, S.P.; Kaverin, N.V.; Amitina, N.N.; Ginevskaya, V.A.

    1991-01-01

    The effect of UV-irradiation on the infectivity of the SAll rotavirus was examined. The time behavior of the inactivation of infectivity generally exhibited the one-hit pattern. The effect was studied with respect to two phenomena, viz. the RNA-protein linkage and the formation of uracil dimers. To determine the number of the latter, purified 3 H-uridine-labelled rotavirus was exposed to UV radiation, and the RNA was extracted and analyzed by paper chromatography in the ascending mode. The formation of photodimers was found to be an important mechanism in the rotavirus inactivation on conventional irradiation, whereas RNA-protein linkages were observed on the application of high doses only. (author). 3 figs., 10 refs

  12. Gamma-radiation effect on diamond and steel during their irradiation in WWER type reactors

    Nikolaenko, V.A.; Karpukhin, V.I.; Amaev, A.D.; Vikhrov, V.I.; Korolev, Yu.N.; Krasikov, E.A.

    1996-01-01

    A study is made into the influence of reactor gamma radiation on expansion of crystal lattice in diamond. The data obtained are compared to those on radiation embrittlement of reactor vessel steels. The necessity of taking into consideration gamma radiation effects on WWER reactor vessel radiation resistance during long-term operation is shown [ru

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

    Allen, T. R.

    1999-01-01

    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 -8 and 5.8 x 10 -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

  14. Role of vanadium carbide traps in reducing the hydrogen embrittlement susceptibility of high strength alloy steels. Final report

    Spencer, G.L.; Duquette, D.J.

    1998-08-01

    High strength alloy steels typically used for gun steel were investigated to determine their susceptibility to hydrogen embrittlement. Although AISI grade 4340 was quite susceptible to hydrogen embrittlement, ASTM A723 steel, which has identical mechanical properties but slightly different chemistries, was not susceptible to hydrogen embrittlement when exposed to the same conditions. The degree of embrittlement was determined by conducting notched tensile testing on uncharged and cathodically charged specimens. Chemical composition was modified to isolate the effect of alloying elements on hydrogen embrittlement susceptibility. Two steels-Modified A723 (C increased from 0.32% to 0.40%) and Modified 4340 (V increased from 0 to O.12%) were tested. X-ray diffraction identified the presence of vanadium carbide, V{sub 4}C{sub 3}, in A-23 steels, and subsequent hydrogen extraction studies evaluated the trapping effect of vanadium carbide. Based on these tests, it was determined that adding vanadium carbide to 4340 significantly decreased hydrogen embrittlement susceptibility because vanadium carbide traps ties up diffusible hydrogen. The effectiveness of these traps is examined and discussed in this paper.

  15. Hydrogen embrittlement of high strength steel electroplated with zincâ  cobalt allo

    Hillier, Elizabeth M. K.; Robinson, M. J.

    2004-01-01

    Slow strain rate tests were performed on quenched and tempered AISI 4340 steel to measure the extent of hydrogen embrittlement caused by electroplating with zincâ  cobalt alloys. The effects of bath composition and pH were studied and compared with results for electrodeposited cadmium and zincâ  10%nickel. It was found that zincâ  1%cobalt alloy coatings caused serious hydrogen embrittlement (EI 0.63); almost as severe as that of cadmium (EI 0.78). Baking cadmium plate...

  16. Hydrogen embrittlement of Zr-2.5Nb PT with temperature

    Oh, Dong Joon; Ahn, Sang Bok; Kim, Young Suk

    2003-01-01

    The aim of this study is to investigate the effect of hydrogen embrittlement of Zr-2.5Nb CANDU pressure tube. The tests were performed at three hydrogen contents for transverse tensile and CCT specimens while the test temperatures were changed (RT to 300 .deg. C). The specimens were directly machined from the tube retaining original curvature using electric discharge machine. Both the transverse tensile and the fracture toughness tests showed the hydrogen embrittlement clearly at RT but this phenomenon was disappeared while the test temperature arrived over 250 .deg. C

  17. Irreversible traps, their influence on the embrittlement of high strength steel

    Mariano, I; Mansilla, G

    2012-01-01

    Hydrogen (H) can be trapped in lattice defects such as vacancies, dislocations, grain boundaries and interfaces between the matrix and precipitates. The effect on the mechanical properties depends on factors inherent in materials such as the activation energy of irreversible traps (H trapped in Network Places) and its sensitivity to embrittlement. Differential scanning calorimetry (DSC) allows the study of those processes in which enthalpy variation occurs. The purpose is to record the difference in enthalpy change that occurs in the sample as a function of temperature or time. This work represents a study of H embrittlement of high strength steel resulfurized

  18. Reduction of helium embrittlement in stainless steel by finely dispersed TiC precipitates

    Kesternich, W.; Rothaut, J.

    1982-01-01

    The He embrittlement effects in two candidate stainless steels for first wall of fusion reactors were studied in creep tests at 700 0 C simulating the He production by He implantation. Creep rupture life before He implantation and reduction of rupture life due to He were superior by orders of magnitude in 1.4970 steel after pertinent pretreatment compared to 316 steel. The high strength and the low He embrittlement result from a fine dispersion of TiC precipitates in the grain interiors. From microstructural investigations a mechanism explaining the high sink efficiency of TiC for He atom accumulation is suggested. (orig.)

  19. Genetic effects of heavy ion irradiation in maize and soybean

    Yatou, Osamu; Amano, Etsuo; Takahashi, Tan.

    1992-01-01

    Somatic mutation on leaves of maize and soybean were observed to investigate genetic effects of heavy ion irradiation. Maize seeds were irradiated with N, Fe and U ions and soybean seeds were irradiated with N ions. This is a preliminary report of the experiment, 1) to examine the mutagenic effects of the heavy ion irradiation, and 2) to evaluate the genetic effects of cosmic ray exposure in a space ship outside the earth. (author)

  20. Late effects of thoracic irradiation in children

    Boelling, T.; Koenemann, S.; Ernst, I.; Willich, N. [Dept. of Radiotherapy, Univ. Hospital of Muenster (Germany)

    2008-06-15

    Purpose: to summarize the literature regarding the late effects of radiotherapy to the thorax in childhood and adolescence with special emphasis on cardiac and pulmonary impairment. Material und methods: the literature was critically reviewed using the PubMed {sup registered} database with the key words 'late effects', 'late sequelae', 'child', 'childhood', 'adolescence', 'radiation', 'radiotherapy', 'thorax', 'lung', 'heart', and 'pulmonary'. Results: 17 publications dealing with radiation-induced pulmonary and cardiac late sequelae in children could be identified and were analyzed in detail. 29 further publications with additional information were also included in the analysis. Pulmonary function impairment after mediastinal irradiation arose in one third of all pediatric patients, even when treatment was performed with normofractionated lower doses (15-25 Gy). Whole lung irradiation was regularly followed by pulmonary function impairment with differing rates in several reports. However, clinically symptomatic function impairment like dyspnea was less frequent. Irradiation of up to 25 Gy (single doses {<=} 2 Gy) to the heart showed little or no cardiac toxicity in analyses of irradiated children (median follow-up 1.3-14.3 years). Doses of > 25 Gy (single doses {<=} 2-3.3 Gy) led to several cardiac dysfunctions. However, new data from adults with longer follow-up may indicate threshold doses as low as 1 Gy. Impairment of skeletal growth, breast hypoplasia, and secondary malignancy were further potential late sequelae. Conclusion: several retrospective reports described radiation-associated late sequelae in children. However, there is still a lack of sufficient data regarding the characterization of dose-volume effects. (orig.)

  1. Effects of irradiation on the vascularity of lung

    Fujiwara, K; Takegawa, Y; Nagase, M; Akiyama, H [Tokushima Univ. (Japan). School of Medicine

    1975-06-01

    Effects of irradiation on the intravascular volume of the lung were studied with respect to changes in intravascular volume over a period of time after irradiation, the effect of fractionation of the dose and the influence of the irradiation dose rate. After a single irradiation with 1000 rad or 3000 rad, applied locally to the lung, the intravascular volume decreased significantly in 1 to 3 months after irradiation. The changes in the intravascular volumes of lungs could be lessened by fractionation of the dose or by low dose rate irradiation.

  2. The effect of microstructural change on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS) irradiated in JOYO/MARICO-1

    Yano, Yasuhide; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro

    2004-03-01

    It is well known that the irradiation embrittlement is one of the most important issues to apply ferritic steels for FBR core materials, although ferritic steels have been considered to be candidate core materials of the commercialized FBR core material because of their superior swelling resistance. In order to evaluate the effects of microstructural changes during irradiation on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS), microstructural observations were performed with transmission electron microscopy on ruptured halves of the half-sized Charpy specimens of PNC-FMS irradiated in the JOYO/MARICO-1. The results obtained in this study are as follows: (1) There was remarkable disappearance of the lath of martensite in the samples irradiated at 650degC, although there was no significant change in microstructures, especially the lath of martensite between the samples irradiated at 500degC and unirradiated. The disappearance of martensitic lath in the samples irradiated at 650degC was larger than that of the samples thermally aged at 650degC. (2) The ductile-brittle transition temperature (DBTT) of irradiated PNC-FMS is judged to increase with the disappearance of martensitic lath and to decrease with the recovery in dislocations. (3) The decrease in the upper shelf energy (USE) of irradiated PNC-FMS is significantly accompanied by the change of precipitation behavior. (4) The Charpy impact properties and microstructures of PNC-FMS irradiated at 500degC were superior under these irradiation conditions. In future, it is necessary to establish how to evaluate Charpy impact properties in a high fluence region, based on theoretical methods introduced from the data gained in low fluence experiments, in addition to expanding the data area widely. (author)

  3. Irradiation effects on plasma diagnostic components

    Nishitani, Takeo; Iida, Toshiyuki; Ikeda, Yujiro

    1998-10-01

    One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m 2 and 1 MWa/m 2 , respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

  4. Neutron irradiation effects in advanced superconductors

    Yoshida, H.; Kodaka, H.; Miyata, K.; Hayashi, Y.; Atobe, K.

    1988-01-01

    This paper reports the effects of neutron irradiation on superconducting transitions studied by susceptibility and resistivity measurements for A15 type compounds, Laves-phase compounds and oxide superconductors. For A15 superconductors, the transition temperature (T c ) decreased with increasing neutron fluence and showed large drop started at about 5 x 10 18 n/cm 2 (E > 0.1 MeV). Post-irradiation annealing gave recovery of T c , but the behaviors were different for the materials with different composition and microstructure. The Laves-phase compounds showed less degradation than the A15 superconductors. For oxide superconductors very sensitive transition change was observed, including the radiation-induced superconductivity

  5. Neutron irradiation effects in pressure vessel steels and weldments

    Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

    1994-12-31

    This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

  6. Honey irradiation: effect on hydroxymethylfurfural content

    Cova, M. C.; Narvaiz, Patricia

    2011-01-01

    Ionizing radiations can be used to inactivate thermo resistant microorganisms in honey, improving its hygienic and sanitary quality. 'American foulbrood' is a disease caused by the sporulated bacteria Paenibacillus larvae larvae which affects bees, diminishes honey production and impairs trade. Clostridium botulinum spores can also be present in honey and cause the disease in very young children. Considering that ionizing radiation can modify molecules, this work was undertaken to evaluate its effect on hydroxymethylfurfural (HMF) content along storage time. HMF is an aldehyde produced as sugars degrade due to excessive heating and/or time, so it is regulated as a freshness indicator in honey. Two varieties of honey, one fluid and the other creamy, of a commercial high quality brand were packaged in polypropylene recipes and stored at room temperature for 14 months. Irradiation was carried out at the semi industrial cobalt-60 facility of the Ezeiza Atomic Centre, about 600,000 Ci of activity, at doses of 0, 10, 20 and 40 kilo Grays, dose rate: 10 kGy/h, dose uniformity:1.1. HMF values, measured according to White ´s spectrophotometric method, increased along storage time in every sample. Irradiation initially diminished HMF content irrespective of dose, trend which was maintained throughout the storage period in creamy honey. Instead, in fluid honey since the fourth month the slopes corresponding to the irradiated samples curves were greater than that of the control one, as a function of dose, rendering higher HMF values in the 40 kGy sample as compared with control since the 10 th storage month. Possibly the prevailing first irradiation effect during storage time is HMF breakdown, followed by enhanced synthesis. (author) [es

  7. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  8. Hydrogen embrittlement and galvanic corrosion of titanium alloys

    Soh, Jeong Ryong; Jeong, Y. H.; Choi, B. K.; Baek, J. H.; Hwang, D. Y.; Choi, B. S.; Lee, D. J

    2000-06-01

    The material properties including the fracture behavior of titanium alloys used as a steam generator tube in SMART can be degraded de to the hydrogen embrittlement and the galvanic corrosion occurring as a result of other materials in contact with titanium alloys in a conducting corrosive environment. In this report the general concepts and trends of hydrogen embrittlement are qualitatively described to adequately understand and expect the fracture behavior from hydrogen within the bulk of materials and under hydrogen containing environments because hydrogen embrittlement may be very complicated process. And the characteristics of galvanic corrosion closely related to hydrogen embrittlement is qualitatively based on wimple electrochemical theory.

  9. Hydrogen embrittlement and galvanic corrosion of titanium alloys

    Soh, Jeong Ryong; Jeong, Y. H.; Choi, B. K.; Baek, J. H.; Hwang, D. Y.; Choi, B. S.; Lee, D. J.

    2000-06-01

    The material properties including the fracture behavior of titanium alloys used as a steam generator tube in SMART can be degraded de to the hydrogen embrittlement and the galvanic corrosion occurring as a result of other materials in contact with titanium alloys in a conducting corrosive environment. In this report the general concepts and trends of hydrogen embrittlement are qualitatively described to adequately understand and expect the fracture behavior from hydrogen within the bulk of materials and under hydrogen containing environments because hydrogen embrittlement may be very complicated process. And the characteristics of galvanic corrosion closely related to hydrogen embrittlement is qualitatively based on wimple electrochemical theory

  10. Effect of irradiation on olfactory function

    Aiba, Tsunemasa; Sugimoto, Midori; Matsuda, Yasuaki; Sugiura, Yoshikazu; Nakai, Yoshiaki; Nakajima, Toshifumi

    1990-01-01

    The effects of therapeutic irradiation on olfactory function were investigated in 20 patients who received radiation therapy because of a malignant tumor of the nose or paranasal sinuses. The standard olfaction test with a T and T olfactometer and an intravenous olfaction test were given before the radiation therapy, during the period of radiation therapy and 1, 3, 6 and 12 months or more later. Five patients whose olfactory epithelium was outside the radiation field showed no damage to olfactory function. The olfactory function of the other 15 patients whose olfactory epithelium had been exposed to radiation was not obviously changed or damaged at the time of radiation therapy. However, 6 months after irradiation, some patients showed a decline in olfactory function, and after 12 months, 4 of 7 patients showed severe damage to olfactory function. These results suggest that a therapeutic dose of irradiation will not cause severe damage to the olfactory function during the period of radiation therapy, but could cause delayed olfactory disorders in some patients after a few years. These olfactory disorders might be caused by damage to or degeneration of the olfactory epithelium or olfactory nerve. (author)

  11. Irradiation effects on polymer-model compounds

    Seguchi, Tadao; Hayakawa, Naohiro; Tamura, Naoyuki; Katsumura, Yosuke; Hayashi, Nariyuki; Tabata, Yoneho

    1991-01-01

    Irradiation effects on n-paraffins and squalane, used as models of polymers, were investigated by product analysis. Four n-paraffins, C 20 H 42 , C 21 H 44 , C 23 H 48 and C 24 H 50 , and squalane (C 30 H 62 ) were γ-irradiated under vacuum in liquid, crystalline and glassy states. The evolved gases were analyzed by gas chromatography and changes in molecular weight were analyzed by liquid chromatography and mass spectroscopy. G-values for crosslinking of n-paraffins were 1.2 for crystalline states (at 25 0 C) and 1.7 for liquid states (at 55 0 C), and showed no difference between odd and even carbon numbers. The G-value of liquid squalane was 1.7; it was 1.3 for the glassy state at low temperature (-77 0 C). Double bonds were common in the crosslinked products, especially after liquid-phase irradiation. The probability of chain scission was estimated as being negligible, though a small number of chain-scission products (which were products of scission at chain-ends or side chains) were observed by gas analysis. (author)

  12. Immunosuppressive effect of total lymphoid irradiation

    Bendel, V.; Medizinische Hochschule Hannover

    1981-01-01

    Contrary to the immunosuppression by means of wholebody irradiation which is known for a long while but connected with considerable side effects and risks, the total lymphoid irradiation (TLI) is a new possibility of immunosuppression the tolerance of which by man is known by virtue of long-standing experiences with the treatment of malignant lymphatic system diseases. In connexion with organ transplantations, TLI might possibly soon be important for the radiotherapeutist. In the experimentation on animals, the unspecific immunosuppression induced by TLI causes a prolonged survival time of allogeneic skin and organ grafts in certain mammals. Furthermore, a formation of blood chimeras combined with specific, permanent tolerance of organ grafts from the bone marrow donor can be caused by bone marrow transplantation after TLI. First experiences with man have been made. In the German literature, TLI has not been mentioned yet. In the present study, a summary is given on the Anglo-Saxon literature, and the first own experiments with regard to the problem of irradiation dose and transplantation interval are presented. (orig.) [de

  13. LYRA and other projects on RPV steel embrittlement study and mitigation of the AMES network

    Debarberis, L.; Estorff, U. von; Crutzen, S.; Beers, M.; Stamm, H.; Vries, M.I. de; Tjoa, G.L.

    1998-01-01

    Within the framework of the European Network AMES, Ageing Materials evaluation and Studies, a number of experimental works on RPV materials embrittlement are carried out at the Institute of Advanced Materials (AIM) of the Joint Research Centre (JRC) of the European Commission (EC). The objectives of AMES are mainly the understanding of the property degradation phenomena of RPV western reference steels like JRQ and HSST, eastern RPV steels like 15X2mFA and 15H2X15, and annealing possibilities. In order to conduct a very high quality irradiation rig, LYRA facility, has been designed and developed at the High Flux Reactor (HFR) Petten. An other dedicated rig, named LIMA, has been developed at the HFR Petten in order to irradiate RPV steels, internals and in-core materials under typical BWR/PWR conditions. The samples can be irradiated in pressurised water up to 160 bar, 320 deg. C, and the water chemistry fully controlled. For irradiation of standard or miniaturised LWR related materials samples, another group of well experienced irradiation devices with inert gas or liquid metals environment are employed. These devices are tailored to their various specific applications. This paper is intended to give information about the structure and the objectives of the existing European network AMES, and to present the various AMES main and spin-off projects, including a brief description on he modelling activities related to RPV materials embrittlement. (author)

  14. Workshop on materials irradiation effects and applications 2012

    Xu, Qiu; Sato, Koichi; Yoshiie, Toshimasa

    2013-01-01

    For the study of the material irradiation effects, irradiation fields with improved control capabilities, advanced post irradiation experiments and well developed data analyses are required. This workshop aims to discuss new results and to plan the future irradiation research in the KUR. General meeting was held from December 14, 2012 to December 15, 2012 with 44 participants and 28 papers were presented. Especially recent experimental results using irradiation facilities in the KUR such as Materials Controlled Irradiation Facility, Low Temperature Loop and LINAC, and results of computer simulation, and fruitful discussions were performed. This volume contains the summary and selected transparencies presented in the meeting. (author)

  15. Significance of rate of work hardening in tempered martensite embrittlement

    Pietikainen, J.

    1995-01-01

    The main explanations for tempered martensite embrittlement are based on the effects of impurities and cementite precipitation on the prior austenite grain boundaries. There are some studies where the rate of work hardening is proposed as a potential reason for the brittleness. One steel was studied by means of a specially developed precision torsional testing device. The test steel had a high Si and Ni content so ε carbide and Fe 3 C appear in quite different tempering temperature ranges. The M S temperature is low enough so that self tempering does not occur. With the testing device it was possible to obtain the true stress - true strain curves to very high deformations. The minimum toughness was always associated with the minimum of rate of work hardening. The change of deformed steel volume before the loss of mechanical stability is proposed as at least one reason for tempered martensite embrittlement. The reasons for the minimum of the rate of work hardening are considered. (orig.)

  16. The biological effectiveness of antiproton irradiation

    Holzscheiter, Michael H.; Bassler, Niels; Agazaryan, Nzhde; Beyer, Gerd; Blackmore, Ewart; DeMarco, John J.; Doser, Michael; Durand, Ralph E.; Hartley, Oliver; Iwamoto, Keisuke S.; Knudsen, Helge V.; Landua, Rolf; Maggiore, Carl; McBride, William H.; Moller, Soren Pape; Petersen, Jorgen; Skarsgard, Lloyd D.; Smathers, James B.; Solberg, Timothy D.; Uggerhoj, Ulrik I.; Vranjes, Sanja; Withers, H. Rodney; Wong, Michelle; Wouters, Bradly G.

    2006-01-01

    Background and purpose: Antiprotons travel through tissue in a manner similar to that for protons until they reach the end of their range where they annihilate and deposit additional energy. This makes them potentially interesting for radiotherapy. The aim of this study was to conduct the first ever measurements of the biological effectiveness of antiprotons. Materials and methods: V79 cells were suspended in a semi-solid matrix and irradiated with 46.7 MeV antiprotons, 48 MeV protons, or 6 Co γ-rays. Clonogenic survival was determined as a function of depth along the particle beams. Dose and particle fluence response relationships were constructed from data in the plateau and Bragg peak regions of the beams and used to assess the biological effectiveness. Results: Due to uncertainties in antiproton dosimetry we defined a new term, called the biologically effective dose ratio (BEDR), which compares the response in a minimally spread out Bragg peak (SOBP) to that in the plateau as a function of particle fluence. This value was ∼3.75 times larger for antiprotons than for protons. This increase arises due to the increased dose deposited in the Bragg peak by annihilation and because this dose has a higher relative biological effectiveness (RBE). Conclusion: We have produced the first measurements of the biological consequences of antiproton irradiation. These data substantiate theoretical predictions of the biological effects of antiproton annihilation within the Bragg peak, and suggest antiprotons warrant further investigation

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

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Alexandreanu, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

  18. Modification of genetic effect of gamma irradiation by electric current

    Grigor'eva, N.N.; Shakhbazov, V.G.

    1985-01-01

    The effect of direct electric current of different value and polarity on genetic sequences of γ-irradiation of Vicia faba seedlings has been studied. The previously found modifying effect of direct electric current is confirmed. The extent and character of this effect depend on the value and polarity of current as well as time between irradiation and electric effects. Current effect modes having no effect on irradiated seedlings protecting cells from injury and the modes aggravating radiation effect have been found. At certain modes the effects of direct electric current on irradiated seedlings changes in the rearrangement spectrum are observed, particularly the number of bridges is increased

  19. Irradiation effects on the gas formation in metals and alloys

    Lucki, G.; Sciani, V.

    1988-08-01

    Gaseous impurities are produced in structural materials and cladding of present fission reactors and of the future fusion reactors. The gases are generated from the following main sources: a) Hydrogen production from the (n, p) nuclear reaction and Helium formation from (n, α) nuclear reaction in the structure of material, in all reactor types. b) Tritium to Helium desintegration in the materials structure and diffusion of plasma fuel, Deuterium and Tritium, of the confinement system the exhaiust system and the breeding blanket fusion reactors. In the present work a review is made emphasyzing the effects of Helium, considered of major influence on the materials embrittlement. Experimental techniques are discussed, as well as, the results obtained in the Radiation damage Div. of the Instituto de Pesquisas Energeticas e Nucleares - CNEN/SP, are analysed. (author) [pt

  20. Phytosanitary irradiation and fresh fruit quality: Cultivar and maturity effects

    Irradiation is an effective quarantine treatment for global trade of fresh produce. Variation in cultivars and maturity stages can impact the tolerance of fresh fruits to irradiation for the purposes of quarantine security. Tolerance thresholds for irradiated fruit are lacking for a large number of ...

  1. The electron irradiation effects in different structures of diodes

    Li Quanfen; Wang Jiaxu

    1993-01-01

    This paper describes the different electron irradiation effects in different structures of diodes and the different results produced by different irradiation ways. From this work, we can know how to choose proper manufacture arts and comprehensive factors according to the structures of diodes and the irradiation conditions

  2. Genetic effects of low-level irradiation

    Selby, P.B.

    1980-01-01

    Recent estimates of the genetic effects of radiation by two widely recognized committees (BEIR III and UNSCEAR 1977) are based to a large extent on data collected in mice using either the specific-locus method or the approach of empirically determining the nature and extent of radiation-induced genetic damage to the skeleton. Both committees made use of doubling-dose and direct methods of estimating genetic hazard. Their estimates can be applied to assessments of risk resulting from medical irradiation in terms both of risk to the population at large and to the individual

  3. Effects of gamma irradiation on antioxidants of medicinal plants

    Jetawattana, Suwimol [The irradiation research for agriculture program, Office of Atoms for Peace, BK (Thailand); Chaichantipyuth, Chaiyo [Faculty of Pharmacy, Chulalongkorn University, BK (Thailand)

    2003-06-01

    The antioxidant effect of water extracts from irradiated medicinal plants on inhibition of lipid peroxidation in human plasma was examined. The results presented herein indicate that crude extracts from 29 kinds, 31 extracts, of medicinal plants, irradiated at 10 and 25 kilo gray. showed no significant change in inhibition of lipid peroxidation in plasma induced by gamma irradiation (p<0.05). It also found that extraction yields in some irradiated plants were increased.

  4. Effects of gamma irradiation on antioxidants of medicinal plants

    Jetawattana, Suwimol; Chaichantipyuth, Chaiyo

    2003-06-01

    The antioxidant effect of water extracts from irradiated medicinal plants on inhibition of lipid peroxidation in human plasma was examined. The results presented herein indicate that crude extracts from 29 kinds, 31 extracts, of medicinal plants, irradiated at 10 and 25 kilo gray. showed no significant change in inhibition of lipid peroxidation in plasma induced by gamma irradiation (p<0.05). It also found that extraction yields in some irradiated plants were increased

  5. Probabilistic approaches applied to damage and embrittlement of structural materials in nuclear power plants

    Vincent, L.

    2012-01-01

    The present study deals with the long-term mechanical behaviour and damage of structural materials in nuclear power plants. An experimental way is first followed to study the thermal fatigue of austenitic stainless steels with a focus on the effects of mean stress and bi-axiality. Furthermore, the measurement of displacement fields by Digital Image Correlation techniques has been successfully used to detect early crack initiation during high cycle fatigue tests. A probabilistic model based on the shielding zones surrounding existing cracks is proposed to describe the development of crack networks. A more numeric way is then followed to study the embrittlement consequences of the irradiation hardening of the bainitic steel constitutive of nuclear pressure vessels. A crystalline plasticity law, developed in agreement with lower scale results (Dislocation Dynamics), is introduced in a Finite Element code in order to run simulations on aggregates and obtain the distributions of the maximum principal stress inside a Representative Volume Element. These distributions are then used to improve the classical Local Approach to Fracture which estimates the probability for a microstructural defect to be loaded up to a critical level. (author) [fr

  6. Effects of irradiation on color and lipid oxidation of prosciutto

    Kong Qiulian; Qi Wenyuan; Yue Ling; Chen Zhijun; Bao Yingzi; Dai Xudong; Xu Yun

    2012-01-01

    This study dealt with the effect of irradiation on the color, irradiation odor and lipid oxidation of prosciutto crudo. The hams were irradiated by γ-ray and electronic beam (EB). Changes of color, irradiation odor, TBA value (TBARS), peroxide value (POV), carbonyl value and conjugated diene value were analyzed and compared with non-irradiated hams. Results showed that color index (a * ) of control, γ-ray irradiated and EB irradiated were 14.39, 9.45 and 11.71 respectively. The ratios of a * /b * were different with the type of rays. The ratio of a * /b * of EB irradiation was same with control, while that of γ-ray irradiation was decreased apparently. γ-ray irradiation had been shown to have apparently detrimental effect on the color and odor of hams, while EB irradiation had little detrimental effect. Irradiation increased POV and conjugated diene value, but the amounts of lipid oxidation products (TBARS, carbonyl value) were less than nonirradiated hams. (authors)

  7. Renal effects of renal x irradiation and induced autoallergic glomerulonephritis

    Rappaport, D.S.; Casarett, G.W.

    1979-01-01

    This study was conducted to determine what influence a single large x-ray exposure of kidney has on the development and course of an experimental autoallergic glomerulonephritis (EAG) in rats. EAG was induced in female Sprague-Dawley rats by immunization with Bordetella pertussis vaccine and homogenate of homologous kidney tissue and Freund's complete adjuvant. Progressive arteriolonephrosclerosis (ANS) was observed in right (irradiated) kidneys following unilateral renal irradiation (1500 rad). Rats were either immunized, sham-immunized, irradiated, sham-irradiated, or both immunized and irradiated. Light and immunofluorescent microscopic observation, urine protein content, and kidney weights were evaluated. In immunized-irradiated animals the effects of irradiation and immunization were largely additive. Immunization did not considerably influence the development and course of ANS and irradiation did not considerably influence the development and course of EAG

  8. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  9. Effects of irradiation on color and lipid oxidation of prosciutto

    Kong Qiulian; Qi Wenyuan; Yue Ling; Chen Zhijun; Bao Yingzi; Dai Xudong; Xu Yun

    2011-01-01

    This study dealt with the effect of irradiation on the color, ordor and lipid oxidation of prosciutto crudo. The hams were irradiated by γ-ray and electronic beam (EB). Changes of color, ordor, TBA value (TBARS), peroxide value (POV), carbonyl value and conjugated diene value were analyzed and compared with nonirradiated hams. Results showed that color index (a * ) of control, γ-ray irradiated and EB irradiated were 14.39, 9.45 and 11.71 respectively. γ-ray irradiation had been shown to have apparently detrimental effect on the color and ordor of hams, while EB irradiation had little detrimental effect. Irradiation increased POV and conjugated diene value, but the amounts of lipid oxidation products (TBARS, carbonyl value) were less than nonirradiated hams. (authors)

  10. Effect of pelvic irradiation of lactose absorption

    Stryker, J.A.; Mortel, R.; Hepner, G.W.

    1978-01-01

    Twenty-four patients undergoing pelvic irradiation for gynecologic malignancies had 14 C-lactose breath tests performed in the first and fifth weeks of their treatment. The 14 C-lactose breath test was performed by administering 2 μCi of 14 C-lactose by mouth along with 50 g of lactose. Breath samples were collected in ethanolic hyamine 1, 2, and 3 hr later; the radioactivity of the trapped 14 CO 2 was determined by liquid scintillation spectroscopy. In the first week of treatment the percentage of administered 14 C excreted as 14 CO 2 at 1, 2, and 3 hr was 1.7 +- 0.8% (mean +- SD), 4.5 +- 1.6%, and 5.8 +- 1.4%, respectively. In the fifth week of treatment the 1-hr, 2-hr, and 3-hr values were 1.2 +- 0.9%, 3.6 +- 2.0%, and 4.7 +- 1.9%, respectively. The difference between the first week and fifth week test results at 1, 2, and 3 hr was statistically significant (t = 2.64, p 14 C-lactose breath test results in the fifth week and the stool frequency at that time (r = -0.44, p 14 C-lactose breath test results in the fifth week were below normal (<1.2%) had nausea at that time. The data suggest that in some patients, lactose malabsorption as a result of the effect of radiation on small intestinal function may be etiologically related to the symptoms of nausea and diarrhea which occur commonly in patients who are undergoing pelvic irradiation. In addition, the results suggest that lactose-containing foods should be restricted in some patients who are undergoing pelvic irradiation to prevent symptoms resulting from radiation-induced lactose intolerance

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

    Okubo, N.; Kondo, K.; Kaji, Y. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Miwa, Y. [Nuclear Energy and Science Directorate, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken (Japan)

    2007-07-01

    Full text of publication follows: Structural materials in fusion reactor with water cooling system will undergo corrosion in aqueous environment and heavier irradiation than that in LWR. Irradiation assisted stress corrosion (IASCC) may be induced in stainless steels exposed in these environment for a long term of reactor operation. The IASCC is considered to be caused in a welding zone. It is difficult to predict and estimate the IASCC, because several irradiation effects (irradiation hardening, swelling, irradiation induced stress relaxation, etc) work intricately. Firstly, effects of residual stress on irradiation hardening were investigated in stainless steels. Specimens used in this study were SUS316 and SUS316L. By bending deformation, the specimens with several % plastic strain, which corresponds to weld residual stress, were prepared. Ion irradiations of 12 MeV Ni{sup 3+} were performed at 330, 400 and 550 deg. C to 45 dpa in TIARA facility at JAEA. No bent specimen was simultaneously irradiated with the bent specimen. The residual stress was estimated by X-ray residual stress measurements before and after the irradiation. The micro-hardness was measured by using nano-indenter. The irradiation hardening and the stress relaxation were changed by irradiation under bending deformation. The residual stress did not relax even for the case of the higher temperature aging at 500 deg. C for the same time of irradiation. The residual stress after ion irradiation, however, relaxed at these experimental temperatures in SUS316L. The hardness was obviously suppressed in bent SUS316L irradiated at 300 deg. C to 6 or 12 dpa. It was evident that irradiation induced stress relaxation occasionally suppressed the irradiation hardening in SUS316L. (authors)

  12. Opening of new field in material science and technology by materials irradiation research

    Kurishita, Hiroaki [Tohoku Univ., Sendai (Japan). Inst. for Materials Research

    1998-03-01

    It is believed that high energy particle irradiation causes severe degradation of materials, and great efforts have been made to reveal the underlying mechanism of such degradation. However, recent progress of the developments of irradiation rigs performed in the Japan Materials Testing Reactor (JMTR) and materials fabrication techniques has enabled to change our understanding of radiation effects on materials from the above pessimistic one to the very challenging one, i.e., irradiation has the beneficial effect of producing new phenomena and/or innovative materials that will not be available without irradiation. An example to be noted is that irradiation with neutrons in JMTR greatly improved the ductility of less ductile metals. This ductility improvement due to irradiation is directly opposite to irradiation embrittlement and is called radiation induced ductilization (RIDU). In this presentation the significance of RIDU and its mechanism will be stated. (author)

  13. Effect of gamma irradiation on Korean traditional multicolored paintwork

    Yoon, Minchul; Kim, Dae-Woon; Choi, Jong-il; Chung, Yong-Jae; Kang, Dai-Ill; Hoon Kim, Gwang; Son, Kwang-Tae; Park, Hae-Jun; Lee, Ju-Woon

    2015-01-01

    Gamma irradiation can destroy fungi and insects involved in the bio-deterioration of organic cultural heritages. However, this irradiation procedure can alter optical and structural properties of historical pigments used in wooden cultural heritage paintings. The crystal structure and color centers of these paintings must be maintained after application of the irradiation procedure. In this study, we investigated the effects of gamma irradiation on Korean traditional multicolored paintwork (Dancheong) for the preservation of wooden cultural heritages. The main pigments in Korean traditional wooden cultural heritages, Sukganju (Hematite; Fe 2 O 3 ), Jangdan (Minium; Pb 3 O 4 ), Whangyun (Crocoite; PbCrO 4 ), and Jidang (Rutile; TiO 2 ), were irradiated by gamma radiation at doses of 1, 5, and 20 kGy. After irradiation, changes in Commision Internationale d’Eclairage (CIE) color values (L*, a*, b*) were measured using the color difference meter, and their structural changes were analyzed using X-ray diffraction (XRD) analysis. The slightly change in less than 1 dE* unit by gamma irradiation was observed, and structural changes in the Dancheong were stable after exposure to 20 kGy gamma irradiation. In addition, gamma irradiation could be applied to painted wooden cultural properties from the Korean Temple. Based on the color values, gamma irradiation of 20 kGy did not affect the Dancheong and stability was maintained for five months. In addition, the fungicidal and insecticidal effect by less than 5 kGy gamma irradiation was conformed. Therefore, the optical and structural properties of Dancheong were maintained after gamma irradiation, which suggested that gamma irradiation can be used for the preservation of wooden cultural heritages painted with Dancheong. - Highlights: • Effects of gamma irradiation on the Dancheong were evaluated. • We confirmed that optical and structural properties of Dancheong were maintained. • Irradiation can contribute the

  14. Irradiation effects on hydrases for biomedical applications

    Furuta, Masakazu; Ohashi, Isao; Oka, Masahito; Hayashi, Toshio

    2000-01-01

    To apply an irradiation technique to sterilize 'Hybrid' biomedical materials including enzymes, we selected papain, a well-characterized plant endopeptidase as a model to examine durability of enzyme activity under the practical irradiation condition in which limited data were available for irradiation inactivation of enzymes. Dry powder and frozen aqueous solution of papain showed significant durability against 60 Co-gamma irradiation suggesting that, the commercial irradiation sterilizing method is applicable without modification. Although irradiation of unfrozen aqueous papain solution showed an unusual change of the enzymatic activity with the increasing doses, and was totally inactivated at 15 kGy, we managed to keep the residual activity more than 50% of initial activity after 30-kGy irradiation, taking such optimum conditions as increasing enzyme concentration from 10 to 100 mg/ml and purging with N 2 gas to suppress the formation of free radicals. (author)

  15. Role of twinning and transformation in hydrogen embrittlement of austenitic stainless steels

    Caskey, G.R. Jr.

    1977-01-01

    Internal hydrogen embrittlement may be viewed as an extreme form of environmental embrittlement that arises following prolonged exposure to a source of hydrogen. Smooth bar tensile specimens of three stainless steels saturated with deuterium (approximately 200 mol D 2 /m 3 ) were pulled to failure in air at 200 to 400 0 K or in liquid nitrogen at 78 0 K. In Type 304L stainless steel and Tenelon ductility losses are a maximum around 200 to 273 0 K; Type 310 stainless steel is not embrittled at this hydrogen concentration. A distinct change in fracture mode accompanies hydrogen embrittlement, with fracture proceeding along coherent boundaries of pre-existing annealing twins. This fracture path is observed in Tenelon at 78 0 K even when hydrogen is absent. There is also a change in fracture appearance in specimens with no prior exposure to hydrogen if they are pulled to failure in high-pressure hydrogen. The fracture path is not identifiable, however. Magnetic response measurements and changes in the stress-strain curves show that hydrogen suppresses formation of strain-induced α'-martensite at 198 0 K in both Type 304L stainless steel and Tenelon, but there is little effect in Type 304L stainless at 273 0 K

  16. Long-term embrittlement of cast duplex stainless steels in LWR systems

    Chopra, O.K.; Chung, H.M.

    1990-08-01

    This progress report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems during the six months from April to September 1988. Characteristics of the primary mechanism of aging embrittlement (i.e., spinodal decomposition of ferrite) and synergistic effects of alloying and impurity elements that influence the kinetics of the primary mechanism are discussed. Several secondary metallurgical processes of embrittlement, strongly dependent on the C, N, Ni, Mo, and Si content of various heats, are identified. Information on kinetics and data on impact properties are analyzed and correlated with microstructural characteristics to provide a unified method of extrapolating accelerated-aging data to reactor operating conditions. Fracture toughness data are presented for several heats of cast stainless steel aged at temperatures between 320 and 450 degrees C for times up to 10,000 h. Mechanical property data are analyzed to develop the procedure and correlations or predicting the kinetics and extent of embrittlement of reactor components from known material parameters. The method and examples of estimating the impact strength and fracture toughness of cast components during reactor service are described. The lower-bound values of impact strength and fracture toughness for cast stainless steels at LWR operating temperatures are defined. 42 refs., 14 figs., 6 tabs

  17. Stress corrosion cracking and hydrogen embrittlement of an Al-Zn-Mg-Cu alloy

    Song, R.G.; Dietzel, W.; Zhang, B.J.; Liu, W.J.; Tseng, M.K.; Atrens, A.

    2004-01-01

    The age hardening, stress corrosion cracking (SCC) and hydrogen embrittlement (HE) of an Al-Zn-Mg-Cu 7175 alloy were investigated experimentally. There were two peak-aged states during ageing. For ageing at 413 K, the strength of the second peak-aged state was slightly higher than that of the first one, whereas the SCC susceptibility was lower, indicating that it is possible to heat treat 7175 to high strength and simultaneously to have high SCC resistance. The SCC susceptibility increased with increasing Mg segregation at the grain boundaries. Hydrogen embrittlement (HE) increased with increased hydrogen charging and decreased with increasing ageing time for the same hydrogen charging conditions. Computer simulations were carried out of (a) the Mg grain boundary segregation using the embedded atom method and (b) the effect of Mg and H segregation on the grain boundary strength using a quasi-chemical approach. The simulations showed that (a) Mg grain boundary segregation in Al-Zn-Mg-Cu alloys is spontaneous, (b) Mg segregation decreases the grain boundary strength, and (c) H embrittles the grain boundary more seriously than does Mg. Therefore, the SCC mechanism of Al-Zn-Mg-Cu alloys is attributed to the combination of HE and Mg segregation induced grain boundary embrittlement

  18. High-temperature helium embrittlement (T>=0,45Tsub(M)) of metals

    Batfalsky, P.

    1984-06-01

    High temperature helium embrittlement, swelling and irradiation creep are the main technical problem of fusion reactor materials. The expected helium production will be very high. The helium produced by (n,α)-processes precipitates into helium bubbles because its solubility in solid metals is very low. Under continuous helium production at high temperature and stress the helium bubbles grow and lead to intergranular early failure. Solution annealed foil specimens of austenitic stainless steel AISI 316 were implanted with α-particles: 1. during creep tests at 1023 K (''in-beam'' test) 2. before the creep tests at high temperature (1023 K). The creep tests have been performed within large ranges of test parameter, e.g. applied stress, temperature, helium implantation rate and helium concentration. After the creep tests the microstructure was investigated using scanning (SEM) and transmission (TEM) electron microscopy. All the helium implanted specimens showed high temperature helium embrittlement, i.e. reduction of rupture time tsub(R) and ductility epsilonsub(R) and evidence of intergranular brittle fracture. The ''in-beam'' creep tests showed greater reduction of rupture time tsub(R) and ductility than the preimplanted creep tests. The comparison of this experimentally obtained data with various theoretical models of high temperature helium embrittlement showed that within the investigated parameter ranges the mechanism controlling the life time of the samples is probably the gas driven stable growth of the helium bubbles within the grain boundaries. (orig.)

  19. PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

    Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

    1990-06-01

    Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs

  20. Fracture toughness prediction for RPV Steels with various degree of embrittlement

    Margolin, B.; Gulenko, A.; Shvetsova, V.

    2003-01-01

    In the present report, predictions of the temperature dependence of cleavage fracture toughness are performed on the basis of the Master Curve approach and a probabilistic model named now the Prometey model. These predictions are performed for reactor pressure vessel steels in different states, the initial (as-produced), irradiated state with moderate degree of embrittlement and in the highly embrittled state. Calculations of the K IC (T) curves may be performed with both approaches on the basis of fracture toughness test results from pre-cracked Charpy specimens at some (one) temperature. The calculated curves are compared with test results. It is shown that the K IC (T) curves for the initial state calculated with the Master Curve approach and the probabilistic model show good agreement. At the same time, for highly embrittled RPV steel, the K IC (T) curve predicted with the Master Curve approach is not an adequate fit to the experimental data, whereas the agreement of the test results and the K IC (T) curve calculated with the probabilistic model is good. An analysis is performed for a possible variation of the K IC (T) curve shape and the scatter in K IC results. (author)

  1. Electron irradiation effects on lithium peroxide

    Kikkawa, Jun; Shiotsuki, Taishi; Shimo, Yusuke; Koshiya, Shogo; Nagai, Takuro; Nito, Takehiro; Kimoto, Koji

    2018-03-01

    In this study, electron irradiation effects on lithium peroxide (Li2O2), which is an important discharge product of Li-air (or Li-O2) batteries, were investigated using selected-area electron diffraction (SAED) and high-energy resolution electron energy-loss spectroscopy (EELS). The results obtained show that Li2O2 to Li2O transformation occurs with 80 and 300 keV incident electrons under high electron dose rates at 20 and -183 °C. The Li2O2 to Li2O transformation rate for 300 keV was 1/5 of that for 80 keV with the irradiation taking place at -183 °C. We also present a series of the EELS spectra that can be used as a criterion to judge the molar ratio of Li2O to Li2O2 in the general systems where Li2O2 and Li2O coexist.

  2. Irradiation effects on plasma diagnostic components

    Nishitani, Takeo [ed.] [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Iida, Toshiyuki; Ikeda, Yujiro [and others

    1998-10-01

    One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m{sup 2} and 1 MWa/m{sup 2}, respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

  3. Genetic effects of feeding irradiated wheat to mice

    Vijayalaxmi

    1976-01-01

    The effects of feeding irradiated wheat in mice on bone marrow and testis chromosomes, germ cell numbers and dominant lethal mutations were investigated. Feeding of freshly irradiated wheat resulted in significantly increased incidence of polyploid cells in bone marrow, aneuploid cells in testis, reduction in number of spermatogonia of types A, B and resting primary spermatocytes as well as a higher mutagenic index. Such a response was not observed when mice were fed stored irradiated wheat. Also there was no difference between the mice fed un-irradiated wheat and stored irradiated wheat. (author)

  4. Effect of electron beam irradiation on fisheries water

    Sarala Selambakkannu; Khomsaton Abu Bakar; Jamaliah Shariff; Suhairi Alimon

    2012-01-01

    This paper studies about water obtained from fish pond of fisheries research centre. Usual water quality parameters such as pH, COD, Turbidity and Ammonia content were analyzed before and after irradiation. Electron beam irradiation was used to irradiate the water with the dose 100 kGy, 200 kGy and 300 kGy. Only high dose was applied on this water as only a limited amount of samples was supplied. All the parameters indicated a slight increase after irradiation except for the ammonia content, which showed a gradual decrease as irradiation dose increases. Sample condition was changed before irradiation in order to obtain more effective results in the following batch. The water sample from fisheries was diluted with distilled water to the ratio of 1:1.This was followed with irradiation at 100 kGy, 200 kGy and 300 kGy. The results still showed an increase in all parameters after irradiation except for ammonia content. For the following irradiation batch, the pH of the sample was adjusted to pH 4 and pH 8 before irradiation. For this sample the irradiation dose selected was only 100 kGy. A higher value of ammonia was observed for the sample with pH 4 after irradiation. Other parameters were almost the same as the first two batches. (author)

  5. Effect of irradiation on functional properties of Gum Tragacanth

    Neda Mollakhalili meybodi

    2017-03-01

    Full Text Available Background and objective: irradiation is a physical treatment in which products are exposed to ionized radiation such as gamma and x rays to improve the security and quality. Hydrocolloids are components that are used in food science to improve texture properties. Exposing to irradiation treatment may change structural and functional properties. By regard to the importance of irradiation on decontaminating of hydrocolloids in food application, the aim of this study is studying the effect of irradiation at different doses on functional properties of Gum Tragacanth in food application. Material and methods: effect of irradiation treatment was studied on the rheological properties, zeta potential, particle size distribution and surface tension of dispersion systems contained 0/5% w/ w gum tragacanth that is irradiated at different doses (0, 0.75. 3, 5 kGy. The effect of irradiation on rheological properties was monitored by rheometer. In order to monitor the effect of irradiation treatment on particle size distribution, zeta potential and surface tension, particle sizer, Brookhaven zeta plus and tensiometer sere used respectively. All treatments were performed three times and the data were analyzed by one way ANOVA. Significant differences between means were identified (P values < 0.05 using Duncan test. Results: Irradiation, change rheologiacal properties and particle size distribution of dispersion contained gum tragacanth. Irradiation treatment up to 0.75 kGy increase zeta potential, but irradiating at higher doses decrease it again. Results of studying parameters showed that irradiation changes the functional properties by affecting on structure. These changes depend on irradiation dose Conclusion: Gum tragacanth irradiation may improve the functional properties by affecting on structure.

  6. Investigation of the effect of some irradiation parameters on the response of various types of dosimeters to electron irradiation

    Farah, K.; Kuntz, F.; Kadri, O.; Ghedira, L.

    2004-01-01

    Several undyed and dyed polymer films are commercially available for dosimetry in intense radiation fields, especially for radiation processing of food and sterilisation of medical devices. The effects of temperature during irradiation and post-irradiation stability, on the response of these dosimeters are of importance to operators of irradiation facilities. The present study investigates the effects of temperature during irradiation by 2.2 MeV electrons beam accelerator and post irradiation storage on the response of several types of dosimeter films. All dosimeters showed a significant effect of temperature during irradiation and post-irradiation storage

  7. Investigation of the effect of some irradiation parameters on the response of various types of dosimeters to electron irradiation

    Farah, K. E-mail: k.farah@cnstn.rnrt.tn; Kuntz, F.; Kadri, O.; Ghedira, L

    2004-10-01

    Several undyed and dyed polymer films are commercially available for dosimetry in intense radiation fields, especially for radiation processing of food and sterilisation of medical devices. The effects of temperature during irradiation and post-irradiation stability, on the response of these dosimeters are of importance to operators of irradiation facilities. The present study investigates the effects of temperature during irradiation by 2.2 MeV electrons beam accelerator and post irradiation storage on the response of several types of dosimeter films. All dosimeters showed a significant effect of temperature during irradiation and post-irradiation storage.

  8. Investigation of moisture-induced embrittlement of iron aluminides. Final report

    Alven, D.A.; Stoloff, N.S. [Rensselaer Polytechnic Inst., Troy, NY (United States). Materials Engineering Dept.

    1997-06-05

    Iron-aluminum alloys with 28 at.% Al and 5 at.% Cr were shown to be susceptible to hydrogen embrittlement by exposure to both gaseous hydrogen and water vapor. This study examined the effect of the addition of zirconium and carbon on the moisture-induced hydrogen embrittlement of an Fe{sub 3}Al,Cr alloy through the evaluation of tensile properties and fatigue crack growth resistance in hydrogen gas and moisture-bearing air. Susceptibility to embrittlement was found to vary with the zirconium content while the carbon addition was found to only affect the fracture toughness. Inherent fatigue crack growth resistance and fracture toughness, as measured in an inert environment, was found to increase with the addition of 0.5 at.% Zr. The combined addition of 0.5 at.% Zr and carbon only increased the fracture toughness. The addition of 1 at.% Zr and carbon was found to have no effect on the crack growth rate when compared to the base alloy. Susceptibility to embrittlement in moisture-bearing environments was found to decrease with the addition of 0.5 at.% Zr. In gaseous hydrogen, the threshold value of the Zr-containing alloys was found to increase above that found in the inert environment while the crack growth resistance was much lower. By varying the frequency of fatigue loading, it was shown that the corrosion fatigue component of the fatigue crack growth rate in an embrittling environment displays a frequency dependence. Hydrogen transport in iron aluminides was shown to occur primarily by a dislocation-assisted transport mechanism. This mechanism, in conjunction with fractography, indicates that the zirconium-containing precipitates act as traps for the hydrogen that is carried along by the dislocations through the lattice.

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

    Chen, Y., E-mail: Yiren_Chen@anl.gov [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Alexandreanu, B.; Chen, W.-Y.; Natesan, K. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Li, Z.; Yang, Y. [University of Florida, Gainesville, FL 32611 (United States); Rao, A.S. [US Nuclear Regulatory Commission, 11545 Rockville Pike, Rockville, MD 20852 (United States)

    2015-11-15

    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.

  10. Kraft cooking of gamma irradiated wood, (1). Effect of alcohol additives on pre-irradiation

    Inaba, M; Meshitsuka, G; Nakano, J [Tokyo Univ. (Japan). Faculty of Agriculture

    1979-12-01

    Studies have been made of kraft cooking of gamma irradiated wood. Beech (Fagus crenata Blume) wood meal suspended in aqueous alkaline alcohol was irradiated up to 1.5 KGy (0.15 Mrad) with gamma rays from a Co-60 source in the presence or absence of oxygen. The irradiated wood meals were washed thoroughly with fresh water, air dried and cooked under the ordinary cooking conditions. The results are summarized as follows: (1) Pre-irradiation in aqueous alkali have negligible effect on kraft cooking. (2) In the case of ethanol addition (50 g/l), pre-irradiation in vacuo shows acceleration of delignification and stabilization of carbohydrates during kraft cooking. Cooked yield gain by pre-irradiation was about 1.2 in all, over the range of delignification from 80 to 90%. Aqueous ethanol without alkali also shows positive but smaller effect than that with alkali. (3) Propanol, iso-propanol and butanol show positive but smaller effects than ethanol. However, methanol does not show any positive effect. (4) Irradiation in the presence of oxygen does not show any attractive effect on kraft cooking.

  11. The effect of irradiation atmosphere on NMR absorption of 60Co γ-rays irradiated polytetrafluoroethylene

    Tutiya, Mituaki

    1975-01-01

    The radiation stability of polytetrafluoroethylene (PTFE) is sensitive to the presence of oxygen during irradiation. In semicrystalline polymers such as PTFE, it is expected that the influence of irradiation differs in crystalline and amorphous regions, therefore the magnitude of irradiation effect in amorphous region was compared with that in crystalline region of PTFE. In the NMR measurement by the author, the main difference between the PTFE samples annealed after irradiation in air and in vacuum was that the increase in crystallinity and the lowering of crystalline transition temperature Tsub(t) were accelerated in the former case. The scission probability of C-C bonds r can be estimated from the radiation-induced increase in crystallinity in the amorphous region of PTFE, and the value of r in the crystalline region can also be calculated. According to previous papers, the ratio of the scission probability in air to that in vacuum was estimated to be about 7 in the amorphous region and 1.8 in the crystalline region of PTFE. From this fact, the following conclusions were drawn. Main chain scission was accelerated when PTFE was irradiated in air. In PTFE, the amorphous region was affected more by irradiation atmosphere than the crystalline region. In both air and vacuum irradiation, the scission probability in the amorphous region was larger than that in the crystalline region. The effect of irradiation on PTFE was affected by oxygen diffusion, the life time of free radicals, and the degree of packing of molecular chains. (Kako, I.)

  12. Effect of irradiation of sensory quality of cigarettes

    Feng Min; Zhu Jiating; Yang Ping; Wang Dening; Gu Guiqiang

    2012-01-01

    3 brands of cigarettes were irradiated and smoked to make sure the effect of irradiation on sensory qualities of cigarettes, luster, aroma, harmony, offensive taste, irritancy, after taste and total scores of irradiated cigarettes were studied. The results showed that, compared with each index of CK, some indexes changed after irradiation, sensory quality of cigarettes might be improved by suitable dose. The sensory qualities of cigarettes of different brands or different styles change differently, though they were irradiated by the same dose. There was no obvious relation between score of any index and irradiation dose, when cigarette of a same brand irradiated by different doses. Above all, the changes of sensory quality of cigarette may meet the requirements of different smokers on the palate. (authors)

  13. Role of radiation embrittlement in reactor vessel integrity assessment

    Marston, T.U.; Chexal, V.K.; Wyckoff, M.

    1982-01-01

    Reactor vessel integrity calculations are complex. The effect of radiation embrittlement on vessel material properties is a very important aspect of any vessel integrity evaluation. The importance of realistic (based on surveillance capsule results) rather than conservative estimates of the material properties (based on regulatory curves) cannot be overestimated. It is also important to make realistic thermal hydraulic and system operations assumptions. In addition, use of actual flaw sizes from in-service inspections (versus hypothetical flaw size selection) will promote realism. Important research results exist that need to be incorporated into the regulatory process. The authors believe results from current research and development efforts will demonstrate that, with reasonable assumptions and best estimate calculations, the safety of even the older reactor vessels with high copper content welds can be assured over their design lifetimes without the need for major fixes. The utilities, through EPRI and the vendors, have dedicated a significant effort to solving the pressurized thermal shock problem

  14. Dose rate effect in food irradiation

    Singh, H.

    1991-08-01

    It has been suggested that the minor losses of nutrients associated with radiation processing may be further reduced by irradiating foods at the high dose rates generally associated with electron beams from accelerators, rather than at the low dose rates typical of gamma irradiation (e.g. 60 Co). This review briefly examines available comparative data on gamma and electron irradiation of foods to evaluate these suggestions. (137 refs., 27 tabs., 11 figs.)

  15. Effects of irradiation of skin flaps

    Sumi, Y.; Ueda, M.; Oka, T.; Torii, S.

    1984-01-01

    The reaction of skin flaps to irradiation and the optimum postoperative time for irradiation was studied in the rat. Flaps showed different reactions depending on the time of irradiation. There was a correlation between the radiosensitivity and the vascularity of the flap. Those flaps in the marginal hypovascular stage of revascularization showed reactions similar to normal skin. However, severe adverse reactions were observed in the marginal hypervascular stage

  16. Effectiveness of irradiation in killing pathogens

    Yeager, J.G.; Ward, R.L.

    1980-01-01

    United States Environmental Protection Agency regulations include gamma ray irradiation of sludge as an approved Process to Further Reduce Pathogens (PFRP) prior to land application. Research at Sandia National Laboratories on pathogen inactivation in sludge by gamma irradiation has demonstrated that the 1 Mrad PFRP dose is capable, by itself, of eliminating bacterial, fungal, and parasitic pathogens from sludge. Gamma irradiation of sludge in conjunction with the required Processes to Significantly Reduce Pathogens (PSRP) should also eliminate the viral hazard from wastewater sludges

  17. Status of pressure vessel embrittlement study in Japan

    Kataoka, Shigeki [Japan Power Engineering and Inspection Corp. (JAPEIC), Chiba (Japan)

    1997-09-01

    The number of nuclear power plants in service for more than 20 years is increasing in Japan. Subsequently, the aging of nuclear power plants will continue to increase and for this reason, the assurance of the safety and reliability of nuclear power plants is becoming more important. Under this circumstances, Japan Government issued a report: ``Specific Concepts in Dealing with Nuclear Power Plant High Aging`` in April, 1996. This report identified that continuous technology development efforts are important to deal with the issues of nuclear power plant aging, and the following items are extracted for important categories to be developed. (1) Aging phenomena evaluation technology. (2) Inspection/monitoring technology (3) Preventive maintenance/repair technology. Japan Power Engineering and Inspection Corporation (JAPEIC) have been implementing various verification test concerning the above items consigned by the Ministry of International Trade and Industry (MITI). This report outlines the Specific Concepts in Dealing with Nuclear Power Plant High Agency and the past achievements and future plans of various verification tests related to irradiation embrittlement of nuclear reactor pressure vessel, mainly related to Pressurized Thermal Shock (PTS). (author). 4 refs, 8 figs, 5 tabs.

  18. Effect of gamma irradiation on drugs

    Crucq, A.S.; Deridder, V.; Engalytcheff, A.; Slegers, C.; Tilquin, P.

    2005-01-01

    Several drugs (ceftazidime, vancomycin, glucagon, erythromycin and dobutamine) were studied in order to determine their radiostability. The methods used to measure the degradation of the drug were the potency and the colour change after irradiation. Electron spin resonance (ESR) is currently being used to detect irradiated foodstuffs and may be a promising technique to detect irradiated drugs. Trapped radicals in cefazolin sodium were studied and quantified by ESR for this purpose. It is proposed that the trapped radicals play an important role in the formation of the final radiolytic compounds. The potency of ceftazidime was not significantly modified after an irradiation of 25 kGy, whereas the potency of erythromycin and dobutamine decreased slightly. Glucagon was revealed to be radiosensitive with a significant decrease in its potency after irradiation. The visible spectra of glucagon and dobutamine did not change significantly after irradiation. The absorbance of erythromycin and vancomycin increased after irradiation. According to European Pharmacopoeia standards, the colour change of ceftazidime is unacceptable. The ESR spectra reveal that the trapped radicals in cefazolin sodium are characteristic of an irradiation. The radical concentration is dependent on the irradiation dose and decays over time. Radical concentration in cefazolin sodium was reduced by 99% after 100 days of storage. These radicals are responsible for about 13% of the measured final radiolytic product. Ionic reactions could also lead to final radiolytic products. (author)

  19. Application of magnetomechanical hysteresis modeling to magnetic techniques for monitoring neutron embrittlement and biaxial stress

    Sablik, M.J.; Kwun, H.; Rollwitz, W.L.; Cadena, D.

    1992-01-01

    The objective is to investigate experimentally and theoretically the effects of neutron embrittlement and biaxial stress on magnetic properties in steels, using various magnetic measurement techniques. Interaction between experiment and modeling should suggest efficient magnetic measurement procedures for determining neutron embrittlement biaxial stress. This should ultimately assist in safety monitoring of nuclear power plants and of gas and oil pipelines. In the first six months of this first year study, magnetic measurements were made on steel surveillance specimens from the Indian Point 2 and D.C. Cook 2 reactors. The specimens previously had been characterized by Charpy tests after specified neutron fluences. Measurements now included: (1) hysteresis loop measurement of coercive force, permeability and remanence, (2) Barkhausen noise amplitude; and (3) higher order nonlinear harmonic analysis of a 1 Hz magnetic excitation. Very good correlation of magnetic parameters with fluence and embrittlement was found for specimens from the Indian Point 2 reactor. The D.C. Cook 2 specimens, however showed poor correlation. Possible contributing factors to this are: (1) metallurgical differences between D.C. Cook 2 and Indian Point 2 specimens; (2) statistical variations in embrittlement parameters for individual samples away from the stated men values; and (3) conversion of the D.C. Cook 2 reactor to a low leakage core configuration in the middle of the period of surveillance. Modeling using a magnetomechanical hysteresis model has begun. The modeling will first focus on why Barkhausen noise and nonlinear harmonic amplitudes appear to be better indicators of embrittlement than the hysteresis loop parameters

  20. Emulation of neutron irradiation effects with protons: validation of principle

    Was, G.S.; Busby, J.T.; Allen, T.; Kenik, E.A.; Jensson, A.; Bruemmer, S.M.; Gan, J.; Edwards, A.D.; Scott, P.M.; Andreson, P.L.

    2002-01-01

    This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 deg. C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 deg. C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 deg. C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a 'W' shaped profile at 1.0 dpa and then into a 'V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary

  1. Hydrogen embrittlement of ASTM A 203 D nuclear structural steel

    Chakravartty, J.K.; Prasad, G.E.; Sinha, T.K.; Asundi, M.K.

    1986-01-01

    The influence of hydrogen on the mechanical properties of ASTM A 203 D nuclear structural steel has been studied by tension, bend and delayed-failure tests at room temperature. While the tension tests of hydrogen charged unnotched specimens reveal no change in ultimate strength and ductility, the effect of hydrogen is manifested in notched specimens (tensile and bend) as a decrease in ultimate strength (maximum load in bend test) and ductility; the effect increases with increasing hydrogen content. It is observed that for a given hydrogen concentration, the decrease in bend ductility is remarkably large compared to that in tensile ductility. Hydrogen charging does not cause any delayed-failure upto 200 h under an applied tensile stress, 0.85 times the notch tensile strength. However delayed failure occurs in hydrogen charged bend samples in less than 10 h under an applied bending load of about 0.80 times of the uncharged maximum load. Fractographs of hydrogen charged unnotched specimens show ductile dimple fracture, while those of notched tension and bend specimens under hydrogen-charged conditions show a mixture of ductile dimple and quasi-cleavage cracking. The proportion of quasi-cleavage cracking increases with increasing hydrogen content and this fracture mode is more predominant in bend specimens. The changes in tensile properties and fracture modes can reasonably be explained by existing theories of hydrogen embrittlement. An attempt is made to explain the significant difference in the embrittlement susceptibility of bend and tensile specimens in the light of difference in triaxiality and plastic zone size near the notch tip. (orig.)

  2. Effect of gamma irradiation on storability of Syrian walnut

    Al-Bachir, M [Atomic Energy Commission, Damascus (Syrian Arab Republic). Dept. of Radiation Technology

    2001-12-01

    Walnut fruits of Baladi variety were irradiated with 0, 0.5, 1.0, 1.5 and 2.0 kGy of gamma irradiation. The irradiated and unirradiated fruits were stored at room temperature (15 to 18 Centigrade) and at a relative humidity of 50 to 70%. Fungal load, proximate composition, chemical changes and sensory properties of nuts were evaluated immediately after irradiation, 6 and 12 months of storage. The results indicated that gamma irradiation reduced the fungal load. Used doses did not cause any significant change in proximate composition of walnuts. Immediately after irradiation, gamma irradiation increased total acidity and decreased iodine value and the volatile basic nitrogen (VBN). whereas, after 12 months of storage, gamma irradiation decreased total acidity and peroxide value and increased iodine value and (VBN). Immediately after irradiation no significant differences were observed between irradiated and non-irradiated samples in flavor and aroma. Whereas, after 12 months of storage higher doses (1.5 and 2.0 kGy) had a negative effect on sensory characteristics. (author)

  3. Effect of gamma irradiation on storability of Syrian walnut

    Al-Bachir, M.

    2002-01-01

    Walnut fruits of Baladi variety were irradiated with 0, 0.5, 1.0, 1.5 and 2.0 kGy of gamma irradiation. The irradiated and unirradiated fruits were stored at room temperature (15 to 18 Centigrade) and at a relative humidity of 50 to 70%. Fungal load, proximate composition, chemical changes and sensory properties of nuts were evaluated immediately after irradiation, 6 and 12 months of storage. The results indicated that gamma irradiation reduced the fungal load. Used doses did not cause any significant change in proximate composition of walnuts. Immediately after irradiation, gamma irradiation increased total acidity and decreased iodine value and the volatile basic nitrogen (VBN). whereas, after 12 months of storage, gamma irradiation decreased total acidity and peroxide value and increased iodine value and (VBN). Immediately after irradiation no significant differences were observed between irradiated and non-irradiated samples in flavor and aroma. Whereas, after 12 months of storage higher doses (1.5 and 2.0 kGy) had a negative effect on sensory characteristics. (author)

  4. Effect of gamma irradiation on the photoluminescence of porous silicon

    Elistratova, M. A., E-mail: Marina.Elistratova@mail.ioffe.ru; Romanov, N. M. [Peter the Great St. Petersburg Polytechnic University (Russian Federation); Goryachev, D. N. [Russian Academy of Sciences, Ioffe Institute (Russian Federation); Zakharova, I. B. [Peter the Great St. Petersburg Polytechnic University (Russian Federation); Sreseli, O. M. [Russian Academy of Sciences, Ioffe Institute (Russian Federation)

    2017-04-15

    The effect of gamma irradiation on the luminescence properties of porous silicon produced by the electrochemical technique is studied. Changes in the photoluminescence intensity between irradiation doses and over a period of several days after the last irradiation are recorded. The quenching of photoluminescence at low irradiation doses and recovery after further irradiation are registered. It is found that porous silicon is strongly oxidized after gamma irradiation and the oxidation process continues for several days after irradiation. It is conceived that the change in the photoluminescence spectra and intensity of porous silicon after gamma irradiation is caused by a change in the passivation type of the porous surface: instead of hydrogen passivation, more stable oxygen passivation is observed. To stabilize the photoluminescence spectra of porous silicon, the use of fullerenes is proposed. No considerable changes in the photoluminescence spectra during irradiation and up to 18 days after irradiation are detected in a porous silicon sample with a thermally deposited fullerene layer. It is shown that porous silicon samples with a deposited C{sub 60} layer are stable to gamma irradiation and oxidation.

  5. Low-temperature embrittlement and fracture of metals with different crystal lattices – Dislocation mechanisms

    V.M. Chernov

    2016-12-01

    Full Text Available The state of a low-temperature embrittlement (cold brittleness and dislocation mechanisms for formation of the temperature of a ductile-brittle transition and brittle fracture of metals (mono- and polycrystals with various crystal lattices (BCC, FCC, HCP are considered. The conditions for their formation connected with a stress-deformed state and strength (low temperature yield strength as well as the fracture breaking stress and mobility of dislocations in the top of a crack of the fractured metal are determined. These conditions can be met for BCC and some HCP metals in the initial state (without irradiation and after a low-temperature damaging (neutron irradiation. These conditions are not met for FCC and many HCP metals. In the process of the damaging (neutron irradiation such conditions are not met also and the state of low-temperature embrittlement of metals is absent (suppressed due to arising various radiation dynamic processes, which increase the mobility of dislocations and worsen the strength characteristics.

  6. Effect of triple ion beam irradiation on mechanical properties of high chromium austenitic stainless steel

    Ioka, Ikuo; Futakawa, Masatoshi; Nanjyo, Yoshiyasu; Kiuchi, Kiyoshi; Anegawa, Takefumi

    2003-01-01

    A high-chromium austenitic stainless steel has been developed for an advanced fuel cladding tube considering waterside corrosion and irradiation embrittlement. The candidate material was irradiated in triple ion (Ni, He, H) beam modes at 573 K up to 50 dpa to simulate irradiation damage by neutron and transmutation product. The change in hardness of the very shallow surface layer of the irradiated specimen was estimated from the slope of load/depth-depth curve which is in direct proportion to the apparent hardness of the specimen. Besides, the Swift's power low constitutive equation (σ=A(ε 0 + ε) n , A: strength coefficient, ε 0 : equivalent strain by cold rolling, n: strain hardening exponent) of the damaged parts was derived from the indentation test combined with an inverse analysis using a finite element method (FEM). For comparison, Type304 stainless steel was investigated as well. Though both Type304SS and candidate material were also hardened by ion irradiation, the increase in apparent hardness of the candidate material was smaller than that of Type304SS. The yield stress and uniform elongation were estimated from the calculated constitutive equation by FEM inverse analysis. The irradiation hardening of the candidate material by irradiation can be expected to be lower than that of Type304SS. (author)

  7. Further application of the cleavage fracture stress model for estimating the T{sub 0} of highly embrittled ferritic steels

    Sreenivasan, P.R.

    2016-02-15

    The semi-empirical cleavage fracture stress model (CFS), based on the microscopic cleavage fracture stress, s{sub f}, for estimating the ASTM E1921 reference temperature (T{sub 0}) of ferritic steels from instrumented impact testing of unprecracked Charpy V-notch specimens is further confirmed by test results for additional steels, including steels highly embrittled by thermal aging or irradiation. In addition to the ferrite-pearlite, bainitic or tempered martensitic steels (which was examined earlier), acicular or polygonal ferrite, precipitation-strengthened or additional simulated heat affected zone steels are also evaluated. The upper limit for the applicability of the present CFS model seems to be T{sub 41J} ∝160 to 170 C or T{sub 0} or T{sub Qcfs} (T{sub 0} estimate from the present CFS model) ∝100 to 120 C. This is not a clear-cut boundary, but indicative of an area of caution where generation and evaluation of further data are required. However, the present work demonstrates the applicability of the present CFS model even to substantially embrittled steels. The earlier doubts expressed about T{sub Qcfs} becoming unduly non-conservative for highly embrittled steels has not been fully substantiated and partly arises from the necessity of modifications in the T{sub 0} evaluation itself at high degrees of embrittlement suggested in the literature.

  8. Effect of brain prenatal irradiations (review)

    Nyagu, A.I.; Loganovskij, K.N.; Loganovskaya, T.K.

    1998-01-01

    Tendency of intellectual deficiency and emotion disturbance among children which were irradiated in womb was found. Study of the risk of endogenic psychic disorder development and, first of all, schizophrenia in pre-natally irradiated children, as a result of Chernobyl catastrophe, is of special interest. 256 refs., 1 tab

  9. Effect of UV laser irradiation on tissue

    Nakayama, Takeyoshi; Kubo, Uichi

    1992-01-01

    Laser-tissue interactions have been investigated through Electron Probe Micro Analysis (EPMA), UV-visible optical absorption and Fourier Transform Infrared Spectroscopy (FTIR). Three excimer lasers, ArF, KrF and XeCl, were used to irradiate tissue; cow thighbone and gelatin thin film. Features of UV laser irradiation are described. (author)

  10. Investigation of helium-induced embrittlement

    Sabelova, V.; Slugen, V.; Krsjak, V.

    2014-01-01

    In this work, the hardness of Fe-9%(wt.) Cr binary alloy implanted by helium ions up to 1000 nm was investigated. The implantations were performed using linear accelerator at temperatures below 80 grad C. Isochronal annealing up to 700 grad C with the step of 100 grad C was applied on the helium implanted samples in order to investigate helium induced embrittlement of material. Obtained results were compared with theoretical calculations of dpa profiles. Due to the results, the nano-hardness technique results to be an appropriate approach to the hardness determination of thin layers of implanted alloys. Both, experimental and theoretical calculation techniques (SRIM) show significant correlation of measured results of induced defects. (authors)

  11. Present status of the disk pressure tests for hydrogen embrittlements

    Fidelle, J.P.

    1988-01-01

    The Disk Pressure Tests (DPT) have been developed considerably. Theoretically: a finite elements mechanical analysis shows the build up of a triaxial stress state already at the beginning of the test, which, with other reasons accounts for the very high sensitivity. Experimentally: for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for environment embrittlement due to H 2 hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably: up to 300 MPa and up to 1000 0 C. Very low strain rate - longer than a month - tests have been able to evidence HGE; of FCC alloys where H diffusivity is low for very oxidation -sensitive metals such as Nb and Ta, effects may appear only at somewhat high rates. The relationship between dynamic tests, static and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analysed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 15O + materials in different conditions. Comparison of HGE tests and service behaviour of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service, which, provided the shape of the stress strain curves is not significantly affected, can be expanded to IHE, HE by other fluids than H 2 , 36 refs

  12. Embrittlement recovery due to annealing of reactor pressure vessel steels

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1996-01-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

  13. Diagnostic experimental results on the hydrogen embrittlement of austenitic steels

    Gavriljuk, V.G.; Shivanyuk, V.N.; Foct, J

    2003-03-14

    Three main available hypotheses of hydrogen embrittlement are analysed in relation to austenitic steels based on the studies of the hydrogen effect on the interatomic bonds, phase transformations and microplastic behaviour. It is shown that hydrogen increases the concentration of free electrons, i.e. enhances the metallic character of atomic interactions, although such a decrease in the interatomic bonding cannot be a reason for brittleness and rather assists an increased plasticity. The hypothesis of the critical role of the hydrogen-induced {epsilon} martensite was tested in the experiment with the hydrogen-charged Si-containing austenitic steel. Both the fraction of the {epsilon} martensite and resistance to hydrogen embrittlement were increased due to Si alloying, which is at variance with the pseudo-hydride hypothesis. The hydrogen-caused early start of the microplastic deformation and an increased mobility of dislocations, which are usually not observed in the common mechanical tests, are revealed by the measurements of the strain-dependent internal friction, which is consistent with the hypothesis of the hydrogen-enhanced localised plasticity. An influence of alloying elements on the enthalpy E{sub H} of hydrogen migration in austenitic steels is studied using the temperature-dependent internal friction and a correlation is found between the values of E{sub H} and hydrogen-caused decrease in plasticity. A mechanism for the transition from the hydrogen-caused microplasticity to the apparent macrobrittle fracture is proposed based on the similarity of the fracture of hydrogenated austenitic steels to that of high nitrogen steels.

  14. Reactivity-flooding effect of the MNSR inner irradiation sites

    Khamis, I.; Khattab, K.

    1999-01-01

    For the purpose of safety assessments, evaluation of the reactivity effects of inner irradiation sites, being flooded with water in the MNSR reactor was conducted both numerically and experimentally. Measured and calculated effect of different combination of inner irradiation sites being flooded with water was evaluated numerically and experimentally. Good agreement between measurement and calculated results were obtained

  15. Embrittlement of zircaloy cladding due to oxygen uptake (CBRTTL)

    Reymann, G.A.

    1979-02-01

    A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes

  16. Irradiation and inhomogeneity effects on ductility and toughness of (ODS)-7 -13Cr steels

    Preininger, D.

    2007-01-01

    Full text of publication follows: The superimposed effect of irradiation defect and structural inhomogeneity formation on tensile ductility and dynamic toughness of ferritic-martensitic 7-13CrW(Mo)VTa(Nb) and oxide dispersion-strengthened (ODS)-7-13CrWVTa(Ti)- RAFM steels has been examined by work hardening and local stress/strain-induced ductile fracture models. Structural inhomogeneities which strongly promoting plastic instability and localized flow might be formed by the applied fabrication process, high dose irradiation and additionally further during deformation by enhanced local dislocation generation around fine particles or due to slip band formation with localized heating at high impact strain rates ε'. The work hardening model takes into account superimposed dislocation multiplication from stored dislocations, dispersions and also grain boundaries as well as annihilation by cross-slip. Analytical relations have been deduced from the model describing uniform ductility and ductile upper shelf energy (USE) observed from Charpy-impact testes. Especially, the influence of different irradiation defects like atomic clusters, dislocation loops and coherent chromium-rich α'- precipitates have been considered together with effects from strain rate as well as irradiation (TI) and test temperature TT. Strengthening by clusters and more pronounced by dislocation loops formed at higher TI>250 deg. C reduces uniform ductility and also distinctly stronger dynamic toughness USE. A superimposed hardening by the α'- formation in higher Cr containing 9-13Cr steels strongly reduces toughness assisted by a combined grain-boundary embrittlement with reduction of the ductile fracture stress. But that improves work hardening and uniform ductility as observed particularly due to nano-scale Y 2 O 3 - dispersions in ODS-RAFM steels. For ODS- steels additionally the strength-induced reduction of toughness is diminished by a combined microstructural-induced increase of the ductile

  17. Effect of gamma irradiation on Korean traditional multicolored paintwork

    Yoon, Minchul; Kim, Dae-Woon; Choi, Jong-il; Chung, Yong-Jae; Kang, Dai-Ill; Hoon Kim, Gwang; Son, Kwang-Tae; Park, Hae-Jun; Lee, Ju-Woon

    2015-10-01

    Gamma irradiation can destroy fungi and insects involved in the bio-deterioration of organic cultural heritages. However, this irradiation procedure can alter optical and structural properties of historical pigments used in wooden cultural heritage paintings. The crystal structure and color centers of these paintings must be maintained after application of the irradiation procedure. In this study, we investigated the effects of gamma irradiation on Korean traditional multicolored paintwork (Dancheong) for the preservation of wooden cultural heritages. The main pigments in Korean traditional wooden cultural heritages, Sukganju (Hematite; Fe2O3), Jangdan (Minium; Pb3O4), Whangyun (Crocoite; PbCrO4), and Jidang (Rutile; TiO2), were irradiated by gamma radiation at doses of 1, 5, and 20 kGy. After irradiation, changes in Commision Internationale d'Eclairage (CIE) color values (L*, a*, b*) were measured using the color difference meter, and their structural changes were analyzed using X-ray diffraction (XRD) analysis. The slightly change in less than 1 dE* unit by gamma irradiation was observed, and structural changes in the Dancheong were stable after exposure to 20 kGy gamma irradiation. In addition, gamma irradiation could be applied to painted wooden cultural properties from the Korean Temple. Based on the color values, gamma irradiation of 20 kGy did not affect the Dancheong and stability was maintained for five months. In addition, the fungicidal and insecticidal effect by less than 5 kGy gamma irradiation was conformed. Therefore, the optical and structural properties of Dancheong were maintained after gamma irradiation, which suggested that gamma irradiation can be used for the preservation of wooden cultural heritages painted with Dancheong.

  18. Induced effect of irradiated exogenous DNA on wheat

    Li Zhongjie; Sun Guangzu; Wang Guangjin

    1996-01-01

    Irradiated exogenous DNA introduced into wheat can give rise to break of DNA-chain and damage of part of alkali radicals. Introducing exogenous DNA irradiated by γ rays could increase Do fructification rate and decrease seed size and plumpness. These tendencies became obvious with dose increase. In comparison with control DNA, introducing DNA irradiated could raise evidently mutagenic effect of pollen tube pathway technique

  19. Irradiation effects on the variability of yield characteristics of soybeans

    Pasztor, K.; Egri, K.; Toeroek, Z.; Bornemiszane, P.P.

    1983-01-01

    The seeds of soybean varieties 'Merit' and 'S-1346' were irradiated by fast neutrons with doses between 4 and 174 Gy. The doses in the range of 57-174 Gy proved to be lethal. After low dose irradiation, shorter breeding time and the stimulation of plant growth could be observed. The effects of irradiation on the oil and protein contents of soybeans were contradictory. (V.N.)

  20. Initial assessment of the mechanisms and significance of low-temperature embrittlement of cast stainless steels in LWR systems

    Chopra, O.K.; Sather, A.

    1990-08-01

    This report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems. Metallurgical characterization and mechanical property data from Charpy-impact, tensile, and J-R curve tests are presented for several experimental and commercial heats, as well as for reactor-aged CF-3, CF-8, and CF-8M cast stainless steels. The effects of material variables on the embrittlement of cast stainless steels are evaluated. Chemical composition and ferrite morphology strongly affect the extent and kinetics of embrittlement. In general, the low-carbon CF-3 stainless steels are the most resistant and the molybdenum-containing high-carbon CF-8M stainless steels are most susceptible to embrittlement. The microstructural and mechanical-property data are analyzed to establish the mechanisms of embrittlement. The procedure and correlations for predicting the impact strength and fracture toughness of cast components during reactor service are described. The lower bound values of impact strength and fracture toughness for low-temperature-aged cast stainless steel are defined. 39 refs., 56 figs., 8 tabs

  1. Effects of cryogenic irradiation on temperature sensors

    Courts, S.S.; Holmes, D.S.

    1996-01-01

    Several types of commercially available cryogenic temperature sensors were calibrated, irradiated at 4.2 K by a gamma or neutron source, and recalibrated in-situ to determine their suitability for thermometry in radiation environments. Comparisons were made between pre- and post-irradiation calibrations with the equivalent temperature shift calculated for each sensor at various temperature in the 4.2 K to 330 K range. Four post-irradiation calibrations were performed with annealing steps performed at 20 K, 80 K, and 330 K. Temperature sensors which were gamma irradiated were given a total dose of 10,000 Gy. Temperature sensors which were neutron irradiated were irradiated to a total fluence of 2 x 10 12 n/cm 2 . In general, for gamma radiation environments, diodes are unsuitable for use. Both carbon glass and germanium resistance sensors performed well at lower temperature, while platinum resistance sensors performed best above 30 K. Thin-film rhodium and Cernox trademark resistance sensors both performed well over the 4.2 K to 330 K range. Only thin-film rhodium and Cernox trademark resistance temperature sensors were neutron irradiated and they both performed well over the 4.2 K to 330 K range

  2. Effects of irradiation on ceramics for fusion-reactor applications

    Porter, D.L.

    1982-12-01

    The purpose of this study, coordinated with efforts of LANL and Grumman Aircraft, was to lay some basic groundwork to study the irradiation effects on the engineering properties of some useful classes of ceramic materials; ANL's efforts were pointed towards multiphase materials (glass ceramics and partially-stabilized zirconias). The materials were irradiated at 400 and 550 0 C to fast (E > 0.1 MeV) neutron fluences of approx. 2 x 10 22 n/cm 2 . Fluorophlogapite mica based glass ceramics (Macor, etc.) were found susceptible to weakening due to void formtion between mica plates. Composition variations within this class of glass ceramics seemed to cause sharp variations in the magnitude of the effect. Lithium silicate glass ceramic (ReX) showed sharp contrasts between the effects of ionization irradiation and displacement damage, neutron irradiation having little effect on the ReX structure while electron irradiation creating lithium silicate vitrification and rapid structural annealing

  3. Effect of electron beam irradiation on seed germination

    Han, Seunghee; Bae, Youngmin [Changwon Univ., Changwon (Korea, Republic of)

    2013-07-01

    Effect of electron beam irradiation on seed germination was investigated in this research. Electron beam of 0.5, 1.0, 1.5 and 2.0 kGy was irradiated to the seeds of lettuce, green onion and cucumber, and the irradiated seeds were incubated at 25 .deg. Cn Nitsch medium solidified with 0.2% Phytagel. Germination percentage and the length of the sprouts were determined after 72 hours. Germination percentage of lettuce seeds was greatly reduced by the irradiation, and that of the green onion and cucumber were moderately reduced or unchanged by the irradiation. Although average length of the lettuce sprouts was reduced severely, that of the green onion and cucumber was unchanged or moderately reduced. Conclusively, electron beam irradiation might be a useful way of disinfecting some plant seeds including green onion and cucumber.

  4. Effect of electron beam irradiation on seed germination

    Han, Seunghee; Bae, Youngmin

    2013-01-01

    Effect of electron beam irradiation on seed germination was investigated in this research. Electron beam of 0.5, 1.0, 1.5 and 2.0 kGy was irradiated to the seeds of lettuce, green onion and cucumber, and the irradiated seeds were incubated at 25 .deg. Cn Nitsch medium solidified with 0.2% Phytagel. Germination percentage and the length of the sprouts were determined after 72 hours. Germination percentage of lettuce seeds was greatly reduced by the irradiation, and that of the green onion and cucumber were moderately reduced or unchanged by the irradiation. Although average length of the lettuce sprouts was reduced severely, that of the green onion and cucumber was unchanged or moderately reduced. Conclusively, electron beam irradiation might be a useful way of disinfecting some plant seeds including green onion and cucumber

  5. Effects of electron beam irradiation on cut flowers and mites

    Dohino, Toshiyuki; Tanabe, Kazuo

    1994-01-01

    Two spotted spider mite, Tetranychus urticae KOCH were irradiated with electron beams (2.5MeV) to develop an alternative quarantine treatment for imported cut flowers. The tolerance of eggs increased with age (1-5-day-old). Immature stages (larva-teleiochrysalis) irradiated at 0.4-0.8kGy increased tolerance with their development. Mated mature females irradiated at 0.4kGy or higher did not produce viable eggs, although temporary recovery was observed at 0.2kGy. Adult males were sterilized at 0.4kGy because non-irradiated virgin females mated with yielded female progeny malformed and sterilized. Various effects of electron beam irradiation were observed when nine species of cut flowers were irradiated in 5MeV Dynamitron accelerator. Chrysanthemum and rose were most sensitive among cut flowers. (author)

  6. Effect of irradiation on biochemistry properties of shrimp allergen

    Gu Kefei; Gao Meixu; Li Chunhong; Li Shurong; Pan Jiarong

    2007-01-01

    Study on the effects of 60 Co γ-rays irradiation at the dose of 0,3,5,7,10 kGy on shrimp allergen biochemistry properties was conducted. The results indicated that the allergen protein molecule can be broken down to smaller molecules or coagulated to larger molecules by irradiation. The hydrophobicity and turbidity of irradiated allergen increased with the increase of absorbed dose. The results also show that allergen solution is more sensitive to irradiation than allergen in solid state or in the whole shrimp. (authors)

  7. Effect of preoperative irradiation on healing of low colorectal anastomoses

    Morgenstern, L.; Sanders, G.; Wahlstrom, E.; Yadegar, J.; Amodeo, P.

    1984-01-01

    The effect of preoperative irradiation on the healing of low colorectal anastomoses was studied experimentally. In 12 dogs in whom preoperative irradiation of 4,000 rads was given before low colorectal stapled anastomosis was performed, anastomotic leakage occurred in 66 percent. More than half of the anastomotic leaks were associated with either severe sepsis or death. In a matched group of control animals that underwent stapled anastomoses without irradiation, no anastomotic complications occurred. The clinical implications of this study are that stapled anastomoses in irradiated colon are at serious risk of anastomotic dehiscence and, therefore, should be protected with a proximal colostomy

  8. Effects of gamma irradiation on antioxidants and ultraviolet stabilizers

    Kawamura, Yoko; Miura, Makiko; Miura, Yukiko; Yamada, Takashi

    1998-01-01

    The effects of gamma irradiation on 18 kinds of antioxidants and 10 kinds of ultraviolet stabilizers, intact or in a polyethylene sheet, were studied. After irradiating at a 30kGy dose, the content of additives themselves did not change and new degradation products were not found. While most antioxidants in polyethylene had a decreased content after irradiation, most ultraviolet stabilizers did not change. During the migration tests with aqueous food simulants, additives were not released from either irradiated or unirradiated sheets. For the migration tests with n-heptane, however, all additives were released from the unirradiated sheet, while most of the antioxidants were not released or released only slightly. (author)

  9. A wide-range embrittlement trend curve for western RPV steels

    Kirk, M.T.

    2011-01-01

    Embrittlement trend curves (ETCs) are used to estimate neutron irradiation embrittlement as a function of both exposure (fluence, flux, temperature, ...) and composition variables. ETCs provide information needed to assess the structural integrity of operating nuclear reactors, and to determine their suitability for continued safe operation. Past efforts on ETC development in the United States have used data drawn from domestic licensees. While this approach has addressed past needs well, future needs such as power up-rates, license extensions to 60 years and beyond, and the use of low copper materials in new reactors produce future operating conditions for the US reactor fleet that may differ from past experience, suggesting that data from sources other than licensee surveillance programs may be needed. In this paper we draw together embrittlement data expressed in terms of ΔT41J and ΔYS from a wide variety of data sources as a first step in examining future embrittlement trends. We develop a 'wide range' ETC based on a collection of over 2500 data. We assess how well this ETC models the whole database, as well as significant data subsets. Comparisons presented herein indicate that a single algebraic model, denoted WR-C(5), represents reasonably well both the trends evident in the data overall as well as trends exhibited by four special data subsets. The WR-C(5) model indicates the existence of trends in high fluence data (Φ > 2-3*10 19 n/cm 2 , E > 1 MeV) that are not as apparent in the US surveillance data due to the limited quantity of ΔT30 data measured at high fluence in this dataset. Additionally, WR-C(5) models well the trends in both test and power reactor data despite the fact it has not term to account for flux. It is suggested that one appropriate use of the WR-C(5) trend curve may include the design irradiation studies to validate or refute the findings presented herein. Additionally, WR-C(5) could be used, along with other information (e.g., other

  10. Irradiation and Post-Irradiation Storage of Chicken: Effects on Fat and Proteins

    Abou-Tarboush, H.M.; Al-Kahtani, H.A.; Abou-Arab, A.A.; Atia, M.; Bajaber, A.S.; Ahmed, M.A.; El-Mojaddidi, M.A.

    1997-01-01

    Chicken were subjected to gamma irradiation doses of 2.5, 5.0, 7.5 and 10.0 KGy and post-irradiation storage of 21 days at 4±2º. The effects on fat and protein of chicken were studied. Rate of formation of total volatile basic-nitrogen was less in irradiated samples particularly in samples treated with 5.0KGy during the entire storage. Fatty acid profiles of chicken lipids were not significantly (P≤ 0.05) affected by irradiation especially at doses of 5.0 KGy. However, irradiation caused a large increase in thiobarbituric acid (TBA) values which continued gradually during storage. Changes in amino acids were minimal. Irradiated and unirradiated samples showed the appearance of protein subunits with molecular weights in the range of 10.0 to 88.0 and 10.0 to 67.0 KD, respectively. No changes were observed in the sarcoplasmic protein but the intensity of bands in all irradiated samples decreased after 21 days of storage

  11. Low-energy irradiation effects of gas cluster ion beams

    Houzumi, Shingo; Takeshima, Keigo; Mochiji, Kozo; Toyoda, Noriaki; Yamada, Isao

    2007-01-01

    A cluster-ion irradiation system with cluster-size selection has been developed to study the effects of the cluster size for surface processes using cluster ions. A permanent magnet with a magnetic field of 1.2 T is installed for size separation of large cluster ions. Trace formations at HOPG surface by the irradiation with size-selected Ar-cluster ions under acceleration energy of 30 keV were investigated by a scanning tunneling microscopy. Generation behavior of the crater-like traces is strongly affected by the number of constituent atoms (cluster size) of the irradiating cluster ion. When the incident cluster ion is composed of 100-3000 atoms, crater-like traces are observed on the irradiated surfaces. In contrast, such traces are not observed at all with the irradiation of the cluster-ions composed of over 5000 atoms. Such the behavior is discussed on the basis of the kinetic energy per constituent atom of the cluster ion. To study GCIB irradiation effects against macromolecule, GCIB was irradiated on DNA molecules absorbed on graphite surface. By the GCIB irradiation, much more DNA molecules was sputtered away as compared with the monomer-ion irradiation. (author)

  12. Effect of neutron irradiation on vitreous carbon

    Kurolenkin, E.I.; Virgil'ev, Yu.S.; Chugunova, T.K.

    1989-01-01

    The change in mass (m), volume (V), specific electric resistance (ρ), coefficient of linear thermal expansion (α), dynamic elasticity modulus (E), and limit of bending strength (σ) of vitreous carbon are studied upon neutron irradiation. Samples for study were two forms of vitreous carbon obtained by hardening thermally reactive polymers at 900-1,000 degree K. Phenol-formaldehyde (bakelite lacquer A, Bakelite A) and furfural-phenol-formaldehyde (FM-2) resin were used. They were irradiated in the experimental water - water VVR-M reactor between 360-1,030 degree K. The maximal neutron flux was 1.65·10 21 neut/cm 2 . Neutron irradiation of vitreous carbon led to its shrinkage and accompanied weakening. Shrinkage and weakening of vitreous carbon was decreased with an increase of treatment and irradiation temperatures

  13. An analysis of food irradiation : genetic effects

    MacPhee, D.; Hall, W.

    1988-01-01

    A series of studies undertaken at the National Institute of Nutrition (NIN) in India in the 1970s reported the occurrence of polyploidy in bone-marrow or peripheral lymphocytes in a number of species, including children, fed on freshly irradiated wheat. Opponents of food irradiation use these studies as evidence that genetic damage is caused by the consumption of irradiated food. This review of those NIN studies and of the attempts to replicate them and of two other relevant studies concludes that the claim that consumption of irradiated food causes genetic damage has not been substantiated. Other researchers have been unable to replicate the NIN studies. Polyploidy appears to be a poor indicator of genetic damage and the NIN results are biologically implausible

  14. Multiscale Modeling of Hydrogen Embrittlement for Multiphase Material

    Al-Jabr, Khalid A.

    2014-01-01

    Hydrogen Embrittlement (HE) is a very common failure mechanism induced crack propagation in materials that are utilized in oil and gas industry structural components and equipment. Considering the prediction of HE behavior, which is suggested

  15. Charles J. McMahon Interfacial Segregation and Embrittlement Symposium

    Vitek, Vaclav

    2003-01-01

    .... McMahon Interfacial Segregation and Embrittlement Symposium: Grain Boundary Segregation and Fracture in Steels was sponsored by ASM International, Materials Science Critical Technology Sector, Structural Materials Division, Materials Processing...

  16. Evaluation of the current status of hydrogen embrittlement and stress-corrosion cracking in steels

    Moody, N.R.

    1981-12-01

    A review of recent studies on hydrogen embrittlement and stress-corrosion cracking in steels shows there are several critical areas where data is either ambiguous, contradictory, or non-existent. A relationship exists between impurity segregation and hydrogen embrittlement effects but it is not known if the impurities sensitize a preferred crack path for hydrogen-induced failure or if impurity and hydrogen effects are additive. Furthermore, grain boundary impurities may enhance susceptibility through interactions with some environments. Some studies show that an increase in grain size increases susceptibility; at least one study shows an opposite effect. Recent work also shows that fracture initiates at different locations for external and internal hydrogen environments. How this influences susceptibility is unknown.

  17. Irradiation effects on plasma diagnostic components (2)

    Nishitani, Takeo; Sugie, Tatsuo

    2002-03-01

    Irradiation tests on a number of diagnostic components under fission neutrons, gamma-rays and 14 MeV neutrons have been carried out as a part of the ITER technology R and D program. UV range transmission losses of a KU-1 quartz were measured during 14 MeV neutron and 60 Co gamma-ray irradiation. Significant transmission losses were observed in the wavelength of 200-300 nm. Five kinds of ITER round robin fibers were irradiated in JMTR and the 60 Co gamma-ray irradiation facility. KS-4V, KU-H2G and F-doped fibers have a rather good radiation hardness, which might be available just outside of the vacuum vessel in ITER. Mica substrate bolometer was irradiated in JMTR up to 0.1 dpa. During the cool down phase of the first cycle all connections went open circuit. The use of gold meanders in the bolometer might be problematic in ITER. The magnetic probes were irradiated in JMTR. Drift of 10 - 40 mVs for 1000s was observed with a digital longterm integrator, however, which might be induced not only by RIEMF but also drift inside the integrator itself. ITER-relevant magnetic coil could be made with MI-cables, whose electric drift for 1000-s integration is less than 0.5 mVs. (author)

  18. Late effects of total body irradiation

    Barrett, A.; Gibson, B.

    1987-01-01

    Late effects of chemo-radiotherapy conditioning before bone marrow transplantation (BMT) are being increasingly recognised in long-term survivors, particularly children. They can be divided into two categories: those affecting hormonal status and those affecting specific organ function. All women treated develop ovarian failure with low levels of β-oestradiol and raised values of follicle-stimulating hormone (FSH) and leutinizing hormone (LH). In males, raised FSH and LH values are found with normal testosterone levels but most patients have azoospermia. In children, puberty is usually but not invariably delayed by treatment but can be induced by appropriate hormone replacement. Compensated hypothyroidism was found in 6/30 children. Growth hormone secretion may be impaired especially if previous cranial irradiation has been given. In children, a reduction in sitting height has been observed. Cataract has occurred in 20% of children between 3 and 6 years after treatment. Two second tumours have been observed. No other major organ toxicities have been encountered. (Auth.)

  19. Irradiation effects test series test IE-1 test results report

    Quapp, W.J.; Allison, C.M.; Farrar, L.C.; Mehner, A.S.

    1977-03-01

    The report describes the results of the first programmatic test in the Nuclear Regulatory Commission Irradiation Effects Test Series. This test (IE-1) used four 0.97m long PWR-type fuel rods fabricated from previously irradiated Saxton fuel. The objectives of this test were to evaluate the effect of fuel pellet density on pellet-cladding interaction during a power ramp and to evaluate the influence of the irradiated state of the fuel and cladding on rod behavior during film boiling operation. Data are presented on the behavior of irradiated fuel rods during steady-state operation, a power ramp, and film boiling operation. The effects of as-fabricated gap size, as-fabricated fuel density, rod power, and power ramp rate on pellet-cladding interaction are discussed. Test data are compared with FRAP-T2 computer model predictions, and comments on the consequences of sustained film boiling operation on irradiated fuel rod behavior are provided

  20. Effects of low-level chronic irradiation on aquatic organisms

    Etoh, H. (National Inst. of Radiological Sciences, Chiba (Japan))

    1980-10-01

    Effects of continual irradiation for a long term on fishes and aquatic invertebrates were outlined. Effects of low-level chronic irradiation on aquatic organisms were less than acute effects induced when the same dose was irradiated once. The radiosensitivity of the genital organ to continual irradiation was high. There was a difference in radiosensitivity of the genital organ between female and male, and the degree of the difference varied according to kinds of animals. In an experiment on continual irradiation of adult killifishes, ova recovered from radiation damage, but spermatozoa did not recover. Incubation rates of eggs obtained from aquatic organisms which lived in water where radioactive sewage flowed into decreased significantly, and the frequency of reverse position of salivary gland chromosomes which were peculiar to exposed organisms increased in larvae of Chironomus tentans.

  1. Effect of gamma irradiation on Hom Tong banana

    1971-01-01

    This report contains research on the use of gamma irradiation to retard the ripening and extend the shelf life of bananas. The major concerns were the effects that irradiation would have on the nutritional content, the organoleptic properties and the pigment of the fruit

  2. Effect of insulin on aldolase turnover in irradiated rat liver

    Komov, V.P.; Kirillova, N.V.; Bekdzhanyan, A.G.

    1984-01-01

    A study was made of the effect of insulin on the rate of biosynthesis, ''half life'', spontaneous decomposition and transport of aldolase in mitochondria of liver and blood plasma of rats, subjected to whole-body X-irradiation. The hormone injected after irradiation was shown to normalize the rate of spontaneous decay and the time of aldolase functioning

  3. Effects of irradiation on freshkeeping of rose cut-flower

    Ding Zengcheng; Li Chuntao; Tang Fei; Xu Hongqing; Shi Dan

    2003-01-01

    Effects of irradiation treatment on the freshkeeping of rose cut-flower were studied. The result showed that respiratory rate, Pro, MDA and colour of rose changed after irradiated with 0, 50, 100, 200 and 300 Gy treatments, and the florescence period was prolonged with 200 and 300 Gy treatments

  4. Effect of autoclave processing and gamma irradiation on apparent ...

    The objective of this study was to investigate the effect of autoclaving and different doses of gamma irradiation on the apparent ileal digestibility of amino acids of cottonseed meal in male broiler breeders. Samples were irradiated in a gamma cell at total doses of 15, 30 and 45 kGy. One package (control) was left at room ...

  5. Irradiation as an effective method of food conservation

    Stachowicz, W.

    1994-01-01

    Irradiation as an effective method for food preservation has been introduced. The worldwide history of radiation methods development has been shown. The state of art of international legislation connected with food irradiation and licensing of that technology in different countries has been reviewed. The list of food products commonly accepted for radiation conservation has also been performed

  6. The effects of different source arrangement on the irradiation efficacy

    Liu Hongyue; Shi Peixin; Lin Yin

    1999-01-01

    The effects of 8 different arrangements with 16 pencil sources on irradiation productivity were studied by using a self-designed computer program. The results showed that the fashion of decentralized arrangement had a higher irradiation productivity than that of centralized in a static and uniform field

  7. Effects of electron beam irradiation on tin dioxide gas sensors

    WINTEC

    sensitivity increases more rapidly under high doses of irra- diation than under low doses of irradiation. The electron beam irradiation effects were simulated and the mecha- nism was discussed. Acknowledgements. The authors gratefully acknowledge financial support from the MOST 973 program, grant no. 2006CB705604 ...

  8. Dependence of hydrogen-induced lattice defects and hydrogen embrittlement of cold-drawn pearlitic steels on hydrogen trap state, temperature, strain rate and hydrogen content

    Doshida, Tomoki; Takai, Kenichi

    2014-01-01

    The effects of the hydrogen state, temperature, strain rate and hydrogen content on hydrogen embrittlement susceptibility and hydrogen-induced lattice defects were evaluated for cold-drawn pearlitic steel that absorbed hydrogen in two trapping states. Firstly, tensile tests were carried out under various conditions to evaluate hydrogen embrittlement susceptibility. The results showed that peak 2 hydrogen, desorbed at temperatures above 200 °C as determined by thermal desorption analysis (TDA), had no significant effect on hydrogen embrittlement susceptibility. In contrast, hydrogen embrittlement susceptibility increased in the presence of peak 1 hydrogen, desorbed from room temperature to 200 °C as determined by TDA, at temperatures higher than −30 °C, at lower strain rates and with higher hydrogen content. Next, the same effects on hydrogen-induced lattice defects were also evaluated by TDA using hydrogen as a probe. Peak 2 hydrogen showed no significant effect on either hydrogen-induced lattice defects or hydrogen embrittlement susceptibility. It was found that hydrogen-induced lattice defects formed under the conditions where hydrogen embrittlement susceptibility increased. This relationship indicates that hydrogen embrittlement susceptibility was higher under the conditions where the formation of hydrogen-induced lattice defects tended to be enhanced. Since hydrogen-induced lattice defects formed by the interaction between hydrogen and strain were annihilated by annealing at a temperature of 200 °C, they were presumably vacancies or vacancy clusters. One of the common atomic-level changes that occur in cold-drawn pearlitic steel showing higher hydrogen embrittlement susceptibility is the formation of vacancies and vacancy clusters

  9. Re-examining reactor vessel embrittlement at Chooz A

    Guilleret, J.-C.

    1988-01-01

    The Chooz A PWR experienced an extended shutdown in 1987/88 following indications that the reactor vessel was embrittling more rapidly than expected. Discrepancies between the expected rate and estimates of the actual rate were not easily explained. The huge body of work done since then to establish safety margins and support restart of the plant should provide a model for the owners of other older PWRs grappling with the embrittlement issue. (author)

  10. Development of small punch tests for ductile-brittle transition temperature measurement of temper embrittled Ni-Cr steels

    Baik, J.M.; Kameda, J.; Buck, O.

    1983-01-01

    Small punch tests were developed to determine the ductile-brittle transition temperature of nickel-chromium (Ni-Cr) steels having various degrees of temper embrittlement and various microstructures. It was found that the small punch test clearly shows the ductile-brittle transition behavior of the temper-embrittled steels. The measured values were compared with those obtained from Charpy impact and uniaxial tensile tests. The effects of punch tip shape, a notch, and the strain rate on the ductile-brittle transition behavior were examined. It was found that the combined use of a notch, high strain rates, and a small punch tip strongly affects the ductile-brittle transition behavior. Considerable variations in the data were observed when the small punch tests were performed on coarse-grained steels. Several factors controlling embrittlement measurements of steels are discussed in terms of brittle fracture mechanisms

  11. Damage dosimetry and embrittlement monitoring of nuclear pressure vessels in real time by magnetic properties measurement. Final report

    Ougouag, A.M.; Stubbins, J.F.; Williams, J.F.; Shong, Wei-Ja.

    1995-04-01

    This program developed a nondestructive technique for gauging the progress of embrittlement of nuclear pressure vessel steels (PVS) by means of monitoring radiation-induced changes in magnetic properties. The technique was developed by running a series of experiments in reactor on typical nuclear pressure vessel steels and weldment material. Following irradiation, changes in magnetic properties were measured and correlated with irradiation dose and with mechanical properties changes, where possible. The changes in magnetic properties were unique to the irradiation environment, and were much larger than those produce by thermal aging in the absence of irradiation. Special techniques for magnetic properties change measurement were developed and complimented by more standard magnetic properties measurement techniques including SQUID measurements. The results of the experiments revealed that magnetic properties were very sensitive to irradiation. Changes in microstructurally-related magnetic properties of as much as 40% were noted after irradiation exposure of as little as 10 17 n/cm 2 (E > 0.1 MeV). The magnetic properties changes plateaued out after doses of around as 10 18 n/cm 2 (E > 0.1 MeV). It is unclear whether further changes would be noted at higher doses which would also be useful for tracking the embrittlement phenomenon. This is recommended for further study. The work supported here resulted in several publications in the open scientific literature

  12. TAREG 2.01/00 Project, ''Validation of neutron embrittlement for VVER 1000 and 440/213 RPVs, with emphasis on integrity assessment''

    Ahlstrand, R.; Margolin, B.; Kostylev, V.; Yurchenko, E.; Akbashev, I.; Piminov, V.; Nikolaev, Y.; Koshkin, V.; Kharshenko, V.; Chyrko, L.; Bukhanov, V.; Comsa, O.

    2012-01-01

    The irradiation embrittlement and integrity of the VVER reactors has been an important issue in many EC supported TACIS and PHARE projects since 1990. In the EC annual program 2000 two TACIS projects (TAREG 2.01/00 and 2.01/03) were approved on the issue in order to improve the neutron irradiation embrittlement databases, elaborate new trend curves for the embrittlement and to assess the integrity of the RPVs (Reactor Pressure Vessel) by analysing PTS transients (Pressurized Thermal Shock) for some selected Russian and Ukrainian VVER 1000 and 440/213 NPPs. In this paper the TAREG 2.01/00 project is briefly described with some details from the twin project 2.01/03, which served as a materials testing project, providing inputs for the 1st project. As a result of the project new trend curves for neutron irradiation embrittlement were elaborated, based on upgraded and more reliable surveillance results databases. The PTS study shows that the integrity of the selected VVER RPVs can be ensured to the end of RPV design life. (author)

  13. Effect of irradiation on analgesia induced by morphine and endorphin

    Kim, Jin Kyu; Lee, Byoung Hun; Hyun, Soung Hee; Chung, Ki Myung [KAERI, Daejeon (Korea, Republic of)

    2003-07-01

    Morphine and endorphin administered intracerebroventricularly (i.c.v.) produce analgesia by activating different descending pain inhibitory systems. Gamma irradiation attenuates the acute analgesic action of i.c.v. injected morphine in mice. This study was done to investigate the effect of-irradiation on the analgesia produced by i.c.v. injected morphine and endorphin in male ICR mice. In one group, mice were exposed to whole-body irradiation at a dose of 5 Gy from a {sup 60}Co source and the analgesic effects were tested 5, 30, 60, 90 and 180 min after irradiation using the acetic acid-induced writhing test. The analgesic effect was produced time-dependently and reached its maximum at 90 min after irradiation. Thus, time was fixed in the following studies. In another group, mice were irradiated with 5 Gy and tested 90 minutes later for analgesia produced by i.c.v. administration of morphine or endorphin. Irradiation significantly potentiated the analgesia produced by endorphin. However, the antinociception produced by morphine was not affected by irradiation. These results support the hypothesis that morphine and endorphin administered supraspinally produce antinocieception by different neuronal mechanisms.

  14. Effect of irradiation on analgesia induced by morphine and endorphin

    Kim, Jin Kyu; Lee, Byoung Hun; Hyun, Soung Hee; Chung, Ki Myung

    2003-01-01

    Morphine and endorphin administered intracerebroventricularly (i.c.v.) produce analgesia by activating different descending pain inhibitory systems. Gamma irradiation attenuates the acute analgesic action of i.c.v. injected morphine in mice. This study was done to investigate the effect of-irradiation on the analgesia produced by i.c.v. injected morphine and endorphin in male ICR mice. In one group, mice were exposed to whole-body irradiation at a dose of 5 Gy from a 60 Co source and the analgesic effects were tested 5, 30, 60, 90 and 180 min after irradiation using the acetic acid-induced writhing test. The analgesic effect was produced time-dependently and reached its maximum at 90 min after irradiation. Thus, time was fixed in the following studies. In another group, mice were irradiated with 5 Gy and tested 90 minutes later for analgesia produced by i.c.v. administration of morphine or endorphin. Irradiation significantly potentiated the analgesia produced by endorphin. However, the antinociception produced by morphine was not affected by irradiation. These results support the hypothesis that morphine and endorphin administered supraspinally produce antinocieception by different neuronal mechanisms

  15. Radioprotective effect of exogenic hypoxia in fractionated irradiation

    Kazymbetov, P.; Yarmonenko, S.P.; Vajnson, A.A.

    1988-01-01

    During the experiments with mice it is established, that exogenic hypoxia protective effect (8%O 2 ), evaluated according to survival rate, decreases at the change from single to fractionated irradiation. Dose change factor (DCF) is equal to 1.55 and 1.22-1.31, respectively. Skin protection using exogenic hypoxia at the local fractionated irradiation is expressed more, than at the fractionated one. DCF is equal to 1.56 and 1.28, respectively. Exogenic hypoxia protection effect in the tumor is expressed rather weakly. DCF at single and fractionated irradiation constitutes 1.03 and 1.07-1.13, respectively. Due to skin preferential protection the therapeutic gain factor at irradiation under the exogenic hypoxia conditions constitutes 1.24 and 1.38-1.46, respectively, at single and fractionated irradiation

  16. Postirradiation examination results for the Irradiation Effects Scoping Test 2

    Mehner, A.S.

    1977-01-01

    The postirradiation examination results are reported for two rods from the second scoping test (IE-ST-2) of the Nuclear Regulatory Commission Irradiation Effects Program. The rods were irradiated in the in-pile test loop of the Power Burst Facility at the Idaho National Engineering Laboratory. Rod IE-005 was fabricated from fresh fuel and cladding previously irradiated in the Saxton Reactor. Rod IE-006, fabricated from fresh fuel and unirradiated cladding, was equipped with six developmental cladding surface thermocouples. The rods were preconditioned, power ramped, and then subjected to film boiling operation. The performance of the rods and the developmental thermocouples are evaluated from the post irradiation examination results. The effects of prior irradiation damage in cladding are discussed in relation to fuel rod behavior during a power ramp and subsequent film boiling operation

  17. Effects on storage life and quality of irradiated mangoes

    Pongsuwan, Dara; Therapawa, Wallapa; Akawassapong, Pakinee; Jattanajet, Jumlong.

    1982-01-01

    Investigations on the effect of irradiation at 50 Krad, hot water treatment at 55 degC 5 min and hot water treatment followed by irradiation were carried out on the mature green Keaw mango to eradicate anthracnose disease development and delay ripening. Before introducing all treatments, mangoes were inoculated by Colletotrichum gloeosporioides. All samples were determined after stored at 10 +- 2 degC and at 85% RH for 3 weeks. No difference in disease control between untreated and irradiated batches, but fruits treated with hot water followed by irradiation were significantly different from untreated ones. All treatments were effective in delaying ripening. Higher dosage of irradiation with a combination of hot water treatment was studied on colour break Pimsen Prure mango and 75 Krad after hot water treatment proved to be promising. Further study is being conducted

  18. Ni/boride interfaces and environmental embrittlement in Ni-based superalloys: A first-principles study

    Sanyal, Suchismita; Waghmare, Umesh V.; Hanlon, Timothy; Hall, Ernest L.

    2011-01-01

    Highlights: ► Fracture strengths of Ni/boride interfaces through first-principles calculations. ► Fracture strengths of Ni/boride interfaces are higher than Ni/Ni 3 Al and NiΣ5 grain boundaries. ► Ni/boride interfaces have higher resistance to O-embrittlement than Ni/Ni 3 Al and NiΣ5 grain boundaries. ► CrMo-borides are more effective than Cr-borides in resisting O-embrittlement. ► Electronegativity differences between alloying elements correlate with fracture strengths. - Abstract: Motivated by the vital role played by boride precipitates in Ni-based superalloys in improving mechanical properties such as creep rupture strength, fatigue crack growth rates and improved resistance towards environmental embrittlement , we estimate fracture strength of Ni/boride interfaces through determination of their work of separation using first-principles simulations. We find that the fracture strength of Ni/boride interfaces is higher than that of other commonly occurring interfaces in Ni-alloys, such as Ni Σ-5 grain boundaries and coherent Ni/Ni 3 Al interfaces, and is less susceptible to oxygen-induced embrittlement. Our calculations show how the presence of Mo in Ni/M 5 B 3 (M = Cr, Mo) interfaces leads to additional reduction in oxygen-induced embrittlement. Through Electron-Localization-Function based analyses, we identify the electronic origins of effects of alloying elements on fracture strengths of these interfaces and observe that chemical interactions stemming from electronegativity differences between different atomic species are responsible for the trends in calculated strengths. Our findings should be useful towards designing Ni-based alloys with higher interfacial strengths and reduced oxygen-induced embrittlement.

  19. Reduction of irradiated tumor cells viability under effect of hyperglycemia

    Meshcherikova, V.V.; Voloshina, E.A.

    1983-01-01

    On Ehrlich carcinoma cells adapted to growth in vivo and in vitro, cellular mechanisms of short-term hyperglycemia effect have been studied. It has been found that SH by itself leads to the loss of viability of a part of cells of ELD solid tumors manifesting during the first 24 hours upon irradiation according to the interphase death type. Tumor cell radiation injuries arising under the effect of irradiation, usually non realized up to the first division, under SH conditions potentiate its injury effect. The phenomena observed explain partially selective injury of tumoral cells in the course of irradiation under SH conditions which testifies to the prospects of its use in clinics

  20. Dosimetry, metallurgical and code needs of the U.S. utilities related to radiation embrittlement of nuclear pressure vessels

    Rahn, F.J.; Marston, T.U.; Ozer, O.; Stahlkopf, K.

    1980-01-01

    Codes and regulation guides in the U.S.A., on performance of pressure vessel are examined. Limiting factors in the analysis and prediction of radiation embrittlement in reactor pressure vessels are: accurate measurement of neutron flux and spectrum in-situ, irradiation rate dependence, environmental conditions influence of flaws annealing, analysis of mechanical tests. The establishment of a self-consistent set of irradiated materials properties data taken at realistic flux rates is required, in conjunction with a careful technique in measuring with a careful technique in measuring the fluence and spectrum at the pressure vessel wall and material test specimen positions

  1. Irradiation effects in hydrated zirconium molybdate

    Fourdrin, C., E-mail: chloe.fourdrin@polytechnique.edu [CEA Saclay, DEN/DANS/DPC/SECR/LSRM, 91 191 Gif-sur-Yvette (France); CEA Saclay, DSM/IRAMIS/SIS2M-UMR 3299/Lrad, 91 191 Gif-sur-Yvette (France); Esnouf, S. [CEA Saclay, DSM/IRAMIS/SIS2M-UMR 3299/Lrad, 91 191 Gif-sur-Yvette (France); Dauvois, V. [CEA Saclay, DEN/DANS/DPC/SECR/LSRM, 91 191 Gif-sur-Yvette (France); Renault, J.-P. [CEA Saclay, DSM/IRAMIS/SIS2M-UMR 3299/Lrad, 91 191 Gif-sur-Yvette (France); Venault, L. [CEN Valrho, DEN/DRCP/SCPS/LC2A, 30 207 Bagnols-sur-Ceze (France); Tabarant, M. [CEA Saclay, DEN/DANS/DPC/LRSI, 91 191 Gif-sur-Yvette (France); Durand, D. [CEA Saclay, DEN/DANS/DPC/SECR/LSRM, 91 191 Gif-sur-Yvette (France); Cheniere, A. [CEA Saclay, DEN/DANS/DPC/LRSI, 91 191 Gif-sur-Yvette (France); Lamouroux-Lucas, C. [CEA Saclay, DEN/DANS/DPC/SECR/LSRM, 91 191 Gif-sur-Yvette (France); Cochin, F. [AREVA NC Tour, AREVA, 92 084 Paris La Defense cedex (France)

    2012-07-15

    Hydrated zirconium molybdate is a precipitate formed during the process of spent nuclear fuel dissolution. In order to study the radiation stability of this material, we performed gamma and electron irradiation in a dose range of 10-100 kGy. XRD patterns showed that the crystalline structure is not affected by irradiation. However, the yellow original sample exhibits a blue-grey color after exposure. The resulting samples were analyzed by means of EPR and diffuse reflectance spectroscopy. Two sites for trapped electrons were evidenced leading to a d{sup 1} configuration responsible for the observed coloration. Moreover, a third defect corresponding to a hole trapped on oxygen was observed after electron irradiation at low temperature.

  2. Sanitizing effects of sewage sludge irradiation treatment

    Zhao Yongfu

    2005-01-01

    A large quantity of pathogenic organisms were found in sewage sludge. An investigation was carried out on the relationship in the chain of sludge-soil-vegetable between the survival of pathogenic organisms and the irradiation dosage. After irradiation with 5-6 kGy, coliform group reduced 3 log cycles, and ascarid ova were completely eliminated with a dose of 1 kGy, making the water matched the standard quality of irrigating water. In the soil applied with irradiated sewage sludge, the total bacteria and coliforms group count reduced to one tenth, and alive ascarid ova was not detected. The coliform group on the Chinese cabbage was extremely low and reached the standard of fresh eating. (authors)

  3. Heavy Ion Irradiation Effects in Zirconium Nitride

    Egeland, G.W.; Bond, G.M.; Valdez, J.A.; Swadener, J.G.; McClellan, K.J.; Maloy, S.A.; Sickafus, K.E.; Oliver, B.

    2004-01-01

    Polycrystalline zirconium nitride (ZrN) samples were irradiated with He + , Kr ++ , and Xe ++ ions to high (>1.10 16 ions/cm 2 ) fluences at ∼100 K. Following ion irradiation, transmission electron microscopy (TEM) and grazing incidence X-ray diffraction (GIXRD) were used to analyze the microstructure and crystal structure of the post-irradiated material. For ion doses equivalent to approximately 200 displacements per atom (dpa), ZrN was found to resist any amorphization transformation, based on TEM observations. At very high displacement damage doses, GIXRD measurements revealed tetragonal splitting of some of the diffraction maxima (maxima which are associated with cubic ZrN prior to irradiation). In addition to TEM and GIXRD, mechanical property changes were characterized using nano-indentation. Nano-indentation revealed no change in elastic modulus of ZrN with increasing ion dose, while the hardness of the irradiated ZrN was found to increase significantly with ion dose. Finally, He + ion implanted ZrN samples were annealed to examine He gas retention properties of ZrN as a function of annealing temperature. He gas release was measured using a residual gas analysis (RGA) spectrometer. RGA measurements were performed on He-implanted ZrN samples and on ZrN samples that had also been irradiated with Xe ++ ions, in order to introduce high levels of displacive radiation damage into the matrix. He evolution studies revealed that ZrN samples with high levels of displacement damage due to Xe implantation, show a lower temperature threshold for He release than do pristine ZrN samples. (authors)

  4. Influence of irradiation conditions on the gamma irradiation effect in polyethylene

    Kacarevic-Popovic, Z.; Gal, O.; Novakovic, L.J.; Secerov, B.

    2002-01-01

    Complete text of publication follows. The radiation cross-linking of polyethylene, due to its high cross-linking yield, has resulted in the radiation technology that has found application in radiation production of heat shrinkable structures and in improvement of mechanical and thermo-physical properties of oriented polyethylene objects. It is observed that the cross-linking efficiency decreases when the irradiation is carried out in the presence of oxygen. In order to estimate the conditions that improve cross-linking efficiency, gamma irradiation effect in two types of polyethylene, irradiated in water and air was investigated. The polyethylene samples used were the low density (LDPE) Lotrene CdF 0302 with 45% crystallinity and the high density (HDPE) Hiplex EHM 6003 with 73% crystallinity. Both kinds of samples, fixed in the Pyrex glass tubes, were simultaneously irradiated with 60 Co gamma rays in distilled water and air, at a doses rate of 9,5 kGy/h (determined by the Fricke dosimeter) at room temperature. Radiation induced oxidative degradation was followed through oxygen containing group formation by the carbonyl group band (1720 cm -1 ) and transvinylene group formation by the band at 966 cm -1 in the infrared spectra. Cross-linking efficiency was determined by gel content using the procedure of the extraction in xylene. The monitored effects of gamma irradiation in water and air point to the conclusion that irradiation in water leads to the lower oxidative degradation and higher cross-linking compared with the effects measured after irradiation in air

  5. Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

    Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

  6. Effects of ion beam irradiation on Oncidium lanceanum orchids

    Zaiton Ahmad; Affrida Abu Hassan

    2006-01-01

    Protocorm-like bodies (PLBs) of an orchid (Oncidium lanceanum) were irradiated using 220 MeV 12 C 5+ ions, accelerated by AVF cyclotron at JAEA, Japan in 2005. Five different doses were applied to the PLBs; 0, 1.0, 2.0, 6.0 and 12.0 Gy. Following irradiation, these PLBs were maintained in cultures for germination and multiplication. Irradiation effects on growth and seedling regeneration patterns as well as molecular characteristics of the in vitro cultures were monitored and recorded. In general, average fresh weights of the irradiated PLBs increased progressively by irradiating the explants at 1.0, 2.0 and reached the maximum at 6.0 Gy. The figure however dropped when the explants were irradiated at 12 Gy. Surprisingly, although the highest average fresh weight was recorded on PLBs irradiated at 6.0 Gy, most of these PLBs were not able to regenerate into complete shoots. On average, after 4 months of irradiation, only 21 seedlings were successfully regenerated from each gram of these PLBs. The highest shoot regeneration was recorded on cultures irradiated at 2.0 Gy in which 102 seedlings were obtained from one gram of the PLBs. Some morphological changes were seen on in vitro plantlets derived from PLBs irradiated at doses of 1.0 and 2.0 Gy. Most of the regenerated seedlings have been transferred to glasshouse for further morphological selection. Molecular analysis showed the presence of DNA polymorphisms among the seedlings from different doses of irradiation. (Author)

  7. Effect of gamma irradiation on storability of strawberry (Fragaria sp)

    Al-Bachir, M; Farah, S [Atomic Energy Commission, Dept. of Agriculture, Damascus (Syrian Arab Republic)

    1998-02-01

    Despite the increased production of strawberry in Syria, the storability and marketability of fruits were not well studied. The objectives of this study were to investigate the effect of gamma irradiation on storability of Senga sengana strawberry produced in Syria and the effect of gamma irradiation on fungal sp. i.e. Botrytis; Penicillium; Rhizopus. The fruits were treated with 1 , 2 and 3 KGy of gamma rays. Treated and untreated fruits were stored at 2 to 4 centigrade and 80 to 90 % relative humidity (RH). In order to investigate their marketability, the fruits where held at room temperature (25 to 30 centigrade). Weight loss, microbial decay, and total loss, juice production, pH, total soluble solids of the juice and organoleptic qualities were evaluated throughout the different storage and marketing periods. The results indicate that gamma irradiation decreased the microbial decay and increased the storability and marketability of fruits by 50 and 100% after using 2 and 3 kGy, respectively. D10 were 1.8 and 2.4 for Botrytis and Rhizopus respectively. One day after irradiation total soluble solids and its pH values were increased. Fourteen days later, irradiated fruits produced more juice with higher pH, but total soluble solids were less. Gamma irradiation did not have an effect on aroma and colour of fruits, whereas, 3 kGy of gamma irradiation had an adverse negative effect on taste. (author)

  8. Effects of irradiation on decontamination and nutrients of dehydrated longan

    Zhu Jiating; Feng Min; Tang Yuxin; Lin Jiabin; Yang Ping; Wang Dening

    2011-01-01

    The dehydrated longan were irradiated at the doses of 2, 4, 6, 8 and 10 kGy and the effects of irradiation on nutrients contents and decontamination were studied. Results showed that dehydrated longan were irradiated at 6 kGy, the number of total bacterial count, mold and coliform bacteria accorded with national standards. There were no significant influence on contents of crude fat, ash, Fe, P, Na and V B2 , but the contents of protein, crude fiber, carbohydrate, moisture, Ca, K and V C of irradiated deghdrated longan were significantly different with control. 6-10 kGy irradiation could meet the commercial demands of dehydrated longan decontamination. (authors)

  9. Silicon/HfO2 interface: Effects of proton irradiation

    Maurya, Savita; Radhakrishna, M.

    2015-01-01

    Substrate oxide interfaces are of paramount importance in deciding the quality of the semiconductor devices. In this work we have studied how 200 keV proton irradiation affects the interface of a 13 nm thick, atomic layer deposited hafnium dioxide on silicon substrate. Pre- and post-irradiation electrical measurements are used to quantify the effect of proton irradiation for varying electrode geometries. Proton irradiation introduces positive charge in the oxide and at the interface of Si/HfO 2 interface. The gate current is not very much affected under positive injection since the induced positive charge is compensated by the injected electrons. Current voltage characteristics under negative bias get affected by the proton irradiation

  10. Irradiation of starches for industrial uses: Chemical and physical effects

    Gonzalez, Maria E.

    1999-01-01

    Corn and cassava starches have been irradiated with gamma doses from 10 to 180 kGy and pastes have been prepared by boiling the starches in water. The rheological properties of the pastes have been determined showing that the 10 kGy dose reduces sharply the viscosity of the aqueous pastes. The solubility of the irradiated starches has been also studied. The cassava starch irradiated with 180 kGy is soluble in boiling water and remains soluble at room temperature. After some considerations on the chemical effects of the irradiation it is concluded that the irradiation technique is suitable to replace the chemical treatment in many industrial applications of the starch. (author)

  11. Effects of heavy particle irradiation on central nervous system

    Nojima, Kumie; Nakadai, Taeko; Khono, Yukio

    2006-01-01

    Effects of low dose heavy particle radiation to central nervous system were studied using human embryonal carcinoma (Ntera2=NT2) and Human neuroblastoma cell (NB1). They exposed to heavy ions and X ray 80% confluent cells in culture bottles. The cells were different type about growth and differentiation in the neuron. The apoptosis profile was measured by AnnexinV-EGFP, PI stained and fluorescence-activated cell sorter (FACS). Memory and learning function of adult mice were studied using water maze test after carbon- or iron-ion irradiation. Memory functions were rapidly decreased after irradiation both ions. Iron -ion group were recovered 20 weeks after irradiation C-ion group were recovered 25 weeks after irradiation. Tier memory were still keep at over 100 weeks after irradiation. (author)

  12. Effect of neutron irradiation on p-type silicon

    Sopko, B.

    1973-01-01

    The possibilities are discussed of silicon isotope reactions with neutrons of all energies. In the reactions, 30 Si is converted to a stable phosphorus isotope forming n-type impurities in silicon. The above reactions proceed as a result of thermal neutron irradiation. An experiment is reported involving irradiation of two p-type silicon single crystals having a specific resistance of 2000 ohm.cm and 5000 to 20 000 ohm.cm, respectively, which changed as a result of irradiation into n-type silicon with a given specific resistance. The specific resistance may be pre-calculated from the concentration of impurities and the time of irradiation. The effects of irradiation on other silicon parameters and thus on the suitability of silicon for the manufacture of semiconductor elements are discussed. (J.K.)

  13. Effects of Irradiation on Insect Host-Parasite Relationship

    Rahalkar, G. W.; Ramakrishnan, V. [Biology Division, Bhabha Atomic Research Centre, Trombay, Bombay (India)

    1968-06-15

    Effects of host irradiation on the development of its parasite were investigated. Females of Bracon brevicomis readily accepted irradiated larvae of tile wax moth (Galleria mellonella) and rice moth (Corcyra cephalonica) for oviposition. However, irradiated wax moth larvae adversely influenced the viability of eggs laid on them and also the survival of the parasite grubs feeding on their bodies. The female grubs were affected more than the males. Rice moth larvae, on the other hand, exerted no significant influence on the viability of parasite eggs, but adversely affected the survival of the grubs. The progeny of parents that had been reared on irradiated larvae also exhibited some developmental changes although grown on non-irradiated host larvae, and these changes were more pronounced when G. mellonella was used as the host insect. (author)

  14. Effects of temperature during the irradiation of calcium carbonate

    Negron M, A.; Camargo R, C.; Ramos B, S. [UNAM, Instituto de Ciencias Nucleares, Circuito Exterior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Gomez V, V. [UNAM, Instituto de Quimica, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Uribe, R. M., E-mail: negron@nucleares.unam.mx [Kent State University, College of Technology, Kent 44240 Ohio (United States)

    2015-10-15

    The gamma irradiation of calcium carbonate at different doses (0 to 309 kGy) and temperature regimes (77 K to 298 K) was carried out to study the effects of irradiation temperature. The changes were followed by EPR spectroscopy. We observed the formation of a composite EPR spectrum even at low radiation doses and temperature. There is a strong effect on the evaluation of the radicals formed as a function of irradiation temperature, probably due to the diffusion in the frozen powder. Response curves show that this system tends to saturate at 10 MGy at 298 K. (Author)

  15. Effects of temperature during the irradiation of calcium carbonate

    Negron M, A.; Camargo R, C.; Ramos B, S.; Gomez V, V.; Uribe, R. M.

    2015-10-01

    The gamma irradiation of calcium carbonate at different doses (0 to 309 kGy) and temperature regimes (77 K to 298 K) was carried out to study the effects of irradiation temperature. The changes were followed by EPR spectroscopy. We observed the formation of a composite EPR spectrum even at low radiation doses and temperature. There is a strong effect on the evaluation of the radicals formed as a function of irradiation temperature, probably due to the diffusion in the frozen powder. Response curves show that this system tends to saturate at 10 MGy at 298 K. (Author)

  16. Biological effect of penetration controlled irradiation with ion beams

    Tanaka, Atsushi; Shimizu, Takashi; Kikuchi, Masahiro; Kobayashi, Yasuhiko; Watanabe, Hiroshi [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment; Yamashita, Takao

    1997-03-01

    To investigate the effect of local irradiation with ion beams on biological systems, technique for penetration controlled irradiation has been established. The range in a target was controlled by changing the distance from beam window in the atmosphere, and could be controlled linearly up to about 31 {mu}m in biological material. In addition, the effects of the penetration controlled irradiations with 1.5 MeV/u C and He ions were examined using tobacco pollen. The increased frequency of leaky pollen produced by ion beams suggests that the efficient pollen envelope damages would be induced at the range-end of ion beams. (author)

  17. Some features of irradiated chitosan and its biological effect

    Hai, Le; Hien, Nguyen Quoc; Luan, Le Quang; Hanh, Truong Thi; Man, Nguyen Tan; Ha, Pham Thi Le; Thuy, Tran Thi [Nuclear Research Institute, VAEC, Dalat (Viet Nam); Yoshii, Fumio; Kume, Tamikazu [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    2001-03-01

    Preparation of chitosan oligomer by radiation degradation was carried out on the gamma Co-60 source. The radiation degradation yield (G{sub d}) of the chitosan was found to be of 1.03. The oligochitosan with 50% of dp>8 fraction was obtained by irradiating the 10% (w/v) chitosan solution in 5% acetic acid at 45 kGy for the chitosan having the initial viscometric average molecular weight, Mv=60,000. Irradiated chitosan showed higher antifungal effect than that of unirradiated one. Furthermore, the irradiated chitosan also showed the growth-promotion effect for plants. (author)

  18. Some features of irradiated chitosan and its biological effect

    Hai, Le; Hien, Nguyen Quoc; Luan, Le Quang; Hanh, Truong Thi; Man, Nguyen Tan; Ha, Pham Thi Le; Thuy, Tran Thi; Yoshii, Fumio; Kume, Tamikazu

    2001-01-01

    Preparation of chitosan oligomer by radiation degradation was carried out on the gamma Co-60 source. The radiation degradation yield (G d ) of the chitosan was found to be of 1.03. The oligochitosan with 50% of dp>8 fraction was obtained by irradiating the 10% (w/v) chitosan solution in 5% acetic acid at 45 kGy for the chitosan having the initial viscometric average molecular weight, Mv=60,000. Irradiated chitosan showed higher antifungal effect than that of unirradiated one. Furthermore, the irradiated chitosan also showed the growth-promotion effect for plants. (author)

  19. Irradiation effects on fuels for space reactors

    Ranken, W.A.; Cronenberg, A.W.

    1984-01-01

    A review of irradiation-induced swelling and gas release experience is presented here for the three principal fuels UO 2 , UC, and UN. The primary advantage of UC and UN over UO 2 is higher thermal conductivity and attendant lower fuel temperature at equivalent pellet diameter and power density, while UO 2 offers the distinct benefit of well-known irradiation performance. Irradiation test results indicate that at equivalent burnup, temperature, and porosity conditions, UC experiences higher swelling than UO 2 or UN. Fission gas swelling becomes important at fuel temperatures above 1320 K for UC, and at somewhat higher temperatures for UO 2 and UN. Evidence exists that at equivalent fuel temperatures and burnups, high density UO 2 and UN experience comparable swelling behavior; however, differences in thermal conductivity influence overall irradiation performance. The low conductivity of UO 2 results in higher thermal gradients which contribute to fuel microcracking and gas release. As a result UO 2 exhibits higher fractional gas release than UN, at least or burnups up to about 3%

  20. Late vascular effects in irradiated mice brain

    Yoshii, Yoshihiko; Maki, Yutaka; Phillips, T.L.

    1982-01-01

    The whole brains of mice were irradiated with 250 kVp X-ray at 120 rad min -1 (1.6 mm Cu HVL, TSD 50 cm) and a histological study was done. The dose range of X-irradiation was from 1300 to 2500 rads. i.e., 1300, 1500, 1750, 2000, and 2500 rads. In the microscopic examination, the mice were killed at the regular postirradiation intervals of between 15 and 20, 31 and 40, 41 and 50, 51 and 60, 61 and 70, 71 and 80, 81 and 90, 139 and 177 weeks. A histological examination was performed by a morphometric estimation of vascular lesion in which the degree of the damage to the arterial system was scored through whole serial brain sections. Necrosis (encephalomalacia), atrophy, cell infiltration, and telangiectatic vascular change of the brain, caused as a result of the fibrinoid necrosis of the large artery were observed. Incidence of the fibrinoid necrosis increased dose dependently between 41 and 87 weeks after irradiation. Mean score of fibrinoid necrosis increased dose dependently approximately 60 weeks after irradiation. It is suggested that scores of large vessel damage do relate to dose at 41 - 87 weeks and can be used to quantify the vessel injury and a fibrinoid necrosis of the large vessels may relate to the incidence of radionecrosis. (author)

  1. Effects of irradiated barley on fattening quail (Coturnix coturnix japonica)

    Dahlhelm, H.

    1999-01-01

    For the feeding experiments reported, barley grains irradiated at doses of 2, 10, and 100 kGy were used as a diet. The results obtained revealed no significant effects in the parameters analysed. (orig./CB) [de

  2. Effects of ion beam irradiation on Oncidium lanceanum

    Zaiton Ahmad; Affrida Abu Hassan; Nurul Aliaa Idris; Mohd Nazir Basiran

    2006-01-01

    Protocorm-like bodies (PLBs) of an orchid (Oncidium lanceanum) were irradiated using 220 MeV 12 C 5+ ion, accelerated by AVF cyclotron at JAEA, Japan in 2005. Five different doses were applied to the PLBs; 0, 1.0, 2.0, 6.0 and 12.0 Gy. Following irradiation, these PLBs were maintained in cultures for germination and multiplication. Irradiation effects on growth and seedling regeneration patterns as well as morphological characteristics of the in vitro cultures were monitored and recorded. In general, average fresh weights of the irradiated PLBs increased progressively by irradiating the explants at 1.0, 2.0 and reached the maximum at 6.0 Gy. The figure however dropped when the explants were irradiated at 12 Gy. Surprisingly, although the highest average fresh weight was recorded on PLBs irradiated at 6.0 Gy, most of these PLBs were not able to regenerate into complete shoots. On average, only 21 seedlings were successfully regenerated from each gram of these PLBs. The highest shoot regeneration was recorded on cultures irradiated at 2.0 Gy in which 102 seedlings were obtained from one gram of the PLBs. Most of the regenerated seedlings have been transferred to glass house for morphological screening. Molecular analysis showed the presence of DNA polymorphisms among the seedlings from different doses

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

    Xian, Yongqiang; Liu, Juan [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); University of Chinese Academy of Science, Beijing 100049 (China); Zhang, Chonghong, E-mail: c.h.zhang@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); Chen, Jiachao [Paul Scherrer Institute, Villigen PSI (Switzerland); Yang, Yitao; Zhang, Liqing; Song, Yin [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China)

    2015-06-15

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

  4. Effect of x-ray irradiated rat fetus mandible

    Han, Chang Gun; You, Dong Soo

    1978-01-01

    The effect of irradiation of x-ray to developing rat mandible in the gestation stage was focused on the study of mandible development and the side effect of x-ray irradiation. The author studied the effect of x-ray irradiation with the gestated rat and their offsprings. 100 rads, 200 rads, 300 rads and 400 rads of x-ray was irradiated in regular order schematically at the lower left abdomen of gestated rat. 18 1/2 days after conception, their offsprings were sacrificed and examined their developing mandible with histological findings. The results were as followed. 1. In the 100-200 rads irradiated rat offsprings, bony trabeclulation was revealed irregular shape. In combine with this finding, osteoblast and fibroblast were appeared shrunken of their nucleus and location of eccentric position. 2. In the 300-400 rads irradiated rat offsprings, decrease of fibroblast and osteoblast appearance in the periosteum were prominently observed and empty lacunae were frequently appeared in their bone matrix. 3. The advent of osteoclast and resorption of cortical bone were appeared in proportion to increasing of x-ray irradiation.

  5. Power reactor embrittlement data base (PR-EDB): Uses in evaluating radiation embrittlement of reactor vessels

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1992-01-01

    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current Codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed, computerized data base. Also, such a data is essential for the evaluation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current compilation contains data from 92 reactors and consists of 175 data points for weld materials (79 different welds) and 395 data points for base materials (110 different base materials). The different types of data that are implemented or planned for this data base are discussed. ''User-friendly'' utility programs have been written to investigate a list of problems using this data base. The utility programs are also used to add and upgrade data, retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in this paper

  6. Effect of prenatal irradiation on total litter birth weight

    Angleton, G.M.; Lee, A.C.

    1981-01-01

    Total litter weight at birth was used as a response variable to study the effects of in utero irradiations on birth weight. Analyses were performed in such a manner as to allow for variations in litter size and environmental temperatures. No effects due to irradiation were noted for exposures given 8 days postcoitus (dpc) and 55 dpc. However, for exposures given 28 dpc, a 5% decrement in birth weight was found for an 80 rad dose

  7. Effect of irradiation on vitamin C and acidity

    Jiravatana, V.

    1971-01-01

    Effect of gamma radiation on the total carbohydrates content of Hom Tong banana stored for various time was discussed. The total carbohydrates of unirradiated banana increased more rapidly during storage than those of irradiated ones. Radiation did not appear to have any effect on the total carbohydrate content in the banana. No significant difference was found on the total carbohydrates between irradiated and unirradiated banana at the same stage of ripeness

  8. Materials for cold neutron sources: Cryogenic and irradiation effects

    Alexander, D.J.

    1990-01-01

    Materials for the construction of cold neutron sources must satisfy a range of demands. The cryogenic temperature and irradiation create a severe environment. Candidate materials are identified and existing cold sources are briefly surveyed to determine which materials may be used. Aluminum- and magnesium-based alloys are the preferred materials. Existing data for the effects of cryogenic temperature and near-ambient irradiation on the mechanical properties of these alloys are briefly reviewed, and the very limited information on the effects of cryogenic irradiation are outlined. Generating mechanical property data under cold source operating conditions is a daunting prospect. It is clear that the cold source material will be degraded by neutron irradiation, and so the cold source must be designed as a brittle vessel. The continued effective operation of many different cold sources at a number of reactors makes it clear that this can be accomplished. 46 refs., 8 figs., 2 tab

  9. The effectiveness of immobilization during prostate irradiation

    Bentel, Gunilla C.; Marks, Lawrence B.; Sherouse, George W.; Spencer, David P.; Anscher, Mitchell S.

    1995-01-01

    Purpose: To evaluate the effect of a hemibody foam cradle on the reproducibility of patient setup during external beam radiation treatment of prostate cancer. Methods and Materials: Between January 1992 and April 1993, 74 patients received external beam radiation treatment to the prostate ± nodes, generally with a four-field box technique. Forty-four of the 74 patients had a custom-made hemibody foam cast used in an attempt to improve setup accuracy. A review of the routine weekly port films was performed following the completion of therapy to determine the reproducibility of patient setup in all 74 patients. The physician's request of an isocenter shift was used as an indicator of reproducibility. Neither the treating technologists nor the physicians knew at the time the films were taken that the port films would be reviewed for setup reproducibility at a later date. The results were compared between the patients treated with (44) and without (30) an immobilization device. Results: In the 44 immobilized patients, 213 routine checks of the isocenter were performed during the 7-week course of radiation therapy. In 17.4% of these instances (37 out of 213), an isocenter shift was requested. This rate is compared to 23.1% (30 out of 130) in the 30 patients who did not have the immobilization device (p < 0.2). There was a statistically significant reduction in isocenter shifts requested in the anterior to posterior direction in the patients who were immobilized, 5.1% (9 out of 175) vs. 12.6% (13 out of 103) (p < 0.05, two tailed chi-square test). There was no significant improvement in the reproducibility of isocenter placement in the cephalad to caudal or right to left directions. Conclusions: This custom-made hemibody foam cradle appears to improve the reproducibility of patient setup during the 7-week course of fractionated external beam irradiation for patients with adenocarcinoma of the prostate. This type of immobilization device is now routinely used in our

  10. Effect of UV irradiation on the early development of silkworm embryos, (2). Development of irradiated eggs

    Kobayashi, Y. (Hokkaido Univ., Sapporo (Japan). Faculty of Agriculture)

    1981-02-01

    The development of silkworm eggs irradiated with UV was compared with that of normal eggs. When the eggs were irradiated with UV from the lateral side immediately after oviposition, development was decelerated, but the germ band was produced. The side of the germ band that was irradiated with UV was abnormal with holes, but the opposite side was hole-free and normal. The normal half of the germ band splits longitudinally, but developed along with the abnormal half to form various malformations. When the eggs were irradiated from the ventral side, the ventral part of the germ band was abnormal at the early stage, the germ band did not concentrate to one place, and produced the half-embryos longitudinally divided by the median line. The UV irradiation at the beginning of the blastoderm stage produced similar results. In the areas irradiated by UV, cleavage nuclei invaded into the surrounding protoplasm, and mitotic figures were observed, but the cell number did not increase even with the advance of development unlike normal cells, whereas the sizes of the cells, their nuclei and nucleoli were enlarged, and intercellular space widened so that the cells were no longer in close contact. The germ band cells produced in the non-irradiated area were normal. The above results suggest that when either the protoplasm or the nucleus of a silkworm egg is damaged by UV, the effect first appears as the inhibition of cell division in the germ band, and as the enlargement of the cell, nucleus and nucleoli. It is presumed that this induces the subsequent inhibition of cell differentiation or abnormalities.

  11. Prenatal effects of ancestral irradiation in inbred mice

    Sprackling, L.E.S.

    1975-01-01

    Mice from 13 inbred strains (S, Z, E, Bab, BaB, BrR, C, K, N, Q, G, CFW, CF1) received continuous cobalt 60 irradiation at low dose rates for varying numbers of consecutive generations. Some Bab and BaB mice had received continuous irradiation for from 24 to 31 generations and the other mice had up to six generations of continuous irradiation in their ancestry. At weaning, the mice were removed from the irradiation room and were mated within strains either to sibs or nonsibs. Ancestral and direct irradiation doses were calculated. The ancestral dose was the effective accumulated dose to the progeny of the mated mice. The direct dose was the amount of irradiation received by any mated female from her conception to her weaning. Each irradiated or control female was scored as fertile or sterile and in utero litter counts were made in pregnant females that were dissected past the tenth day of pregnancy; the sum of moles, dead embryos, and live embryos was the total in utero litter size. A ratio of the living embryos to the total number of embryos in utero was determined for each litter. An increase in ancestral or direct irradiation dose significantly decreased fertility in 11 of the 13 strains. The fertility curves for the pooled data were sigmoid in the area of the doses below those that caused complete sterility. Among the controls, there were significant strain differences in total litter size and in the ratio. Strain X--Y plots, with ancestral or direct doses plotted against total litter size or ratio, revealed the tendency for litter size to decrease as dose increased. The only trend shown for ratio was for the litters with ratios of 0.50 or less to appear more frequently among the irradiated mice. The few corpora lutea counts revealed nothing of significance. Generally, there was a definite trend toward fewer mice alive in utero among the irradiated mice

  12. Effects of irradiation on the components of implantable pacemakers

    Kawamura, Shinji; Ono, Seiji; Kuga, Noriyuki; Shiba, Tooru; Hirose, Tetsuo; Matoba, Masaru

    2003-01-01

    The purpose of this study was to examine the effects of irradiation on implantable pacemaker components. The pacemaker was divided into three components: lead wire and electrode, battery, and electrical circuit, and each component was irradiated by X-ray and electron beams, respectively. The pacemaker parameters were measured by both telemetry data of the programmer and directly measured data from the output terminal. The following results were obtained. For the lead wire and electrode, there was no effect on the pacemaker function due to irradiation by X-ray and electron beams. In the case of battery irradiation, there was no change in battery voltage or current up to 236 Gy X-ray dose. In the electrical circuit, the pacemaker reverted to the regular beating rate (fixed-rate mode) immediately after the start of X-ray irradiation, and it continued in this mode during irradiation. In patients with their own heartbeat rhythm, changing to the fixed-rate mode may cause dangerous conditions such as ventricular fibrillation. When the accumulated irradiation dose is increased, another failure can be seen in the output voltage of the pacemaker. The pacing output voltage dropped rapidly by about 40% at 30-88 Gy. Decreasing the output voltage results in pacing disorders, and heart failure may occur. In the telemetry data of the programmer, no change in output voltage could be detected, highlighting the difference between telemetry data and actual pacing data. (author)

  13. Irradiation and storage effect on some characteristics of soy seeds

    Ramirez Ascheri, Diego Palmiro; Devilla, Ivano Alessandro

    2008-01-01

    The irradiation has been applied frequently in seeds conservation to obtain reduction of losses caused by physiologic processes, besides reducing the microbial load. However, the irradiation process for X-rays is not a common practice in seeds; for that, it is necessary to study that process, in order to know irradiation effect on the soy seeds quality. The objective of this work was to verify the irradiation effect of X-rays and the storage period on the water, oil and protein contends of soy seeds (Glycine max L.) variety Emgopa 302. The experiment was represented by the combination of two factors: X-rays dosage with four irradiation levels [0, 50, 65 and 70 kV] and storage period in laboratory atmosphere with five levels (0, 15, 30, 45 and 60 days). The water content, oil and protein in seeds were assayed in the beginning and every 15 days of storage. The results showed the seeds quality stayed unaffected in irradiation function in the beginning of the experience, with alterations after 15 days, was verified a quality decreasing with the increasing of the X-rays dosage. The soy seeds irradiated had reduced quality in elapsing of the storage period. (author)

  14. Effects of irradiation of sewage sludge on heavy metal bioavailability

    Sheppard, S.C.; Mayoh, K.R.

    1986-10-01

    Sewage sludges are a valuable resource to agriculture, but their use is limited by the hazards of pathogens, toxic chemicals and heavy metals. Irradiation can control the pathogens and deactivate some of the toxic chemicals. The relative cost of industrial-scale irradiation using accelerators has decreased progressively. This, coupled with the increasing necessity to recycle wastes, has led to renewed interest in irradiation of sludges. In response to this renewed interest, this report examines what is known about the effects of irradiation on the bioavailability of heavy metals. Very few studies have addressed this topic, although workers in the U.S. have claimed decreased solubility of metals in irradiated sludges. We have also briefly reviewed the general literature on sludge to gain indirect evidence on the likely effects. The scant data, often based on less than ideal experimental methodologies, show no major consistent effects of irradiation on the availability of heavy metals from sludge. The data are not sufficient to rule out such effects entirely, but the effects appear to be fairly subtle and not likely to persist beyond one growth season. 85 refs

  15. Gamma irradiation effects in optical fibres, splitters, and connectors

    Srećković Milesa Ž.

    2017-01-01

    Full Text Available The paper presents a brief overview of contemporary ELION techniques with stress on their use for material modification and dosimetry. In the attempt to avoid some common misjudges of irradiation effects, special attention is paid to exact definition of irradiation geometry and careful adjustment of dose rates, which enable a proper elaboration of experimental results. In particular, effects of g-rays irradiation on properties of commercial optical fibres, splitters, connectors, and fibre joints are examined, which enables monitoring of irradiation effects in complex configurations made of materials with different radiation hardness (resistance. It has been established that g-rays irradiation of the investigated elements influences, in different ways, the transmission of laser beam signals of various wavelengths, under different modulation regimes. After irradiation, the signal attenuation is noticeably larger, both in optical connectors and optical splitter, than before it, and the effect increases in time. The effects are more pronounced at the 99 % than at the 1 % Y-splitter output at both measured wavelengths, and are more pronounced at 1310 nm than at 1550 nm. [Project of the Serbian Ministry of Education, Science and Technological Development, Grant no. III43009 and Grant no. III45012

  16. Radiation irradiation effects: knowledge and doubts

    Nenot, J.C.

    1994-01-01

    People are ever submitted to natural ionizing radiations. But nuclear power plant workers have this natural exposure and one complementary irradiation, in case of nuclear installation serious accident. In normal running, people, people, particularly those in neighbourhood of the installation, are exposed permanently to liquid and gaseous disposal. In case of accident, exposure increases mainly according to usual state. Only the accident states may induce for people or one worker exposure to high dose ionizing radiations. 1 tab

  17. Long-term aging embrittlement of cast duplex stainless steels in LWR systems

    Chopra, O.K.; Chung, H.M.

    1991-01-01

    The primary objectives of this program are to investigate the significance of in-service embrittlement of cast duplex stainless steels in light water reactor (LWR) systems and to evaluate possible remedies for the embrittlement problem in existing and future plants. The scope of the investigation includes three goals: (1) develop a methodology and correlations for predicting the toughness loss suffered by cast stainless steel components during normal and extended life of LWRs, (2) validate the simulation of in-reactor degradation by accelerated aging, and (3) establish the effects of key compositional and metallurgical variables on the kinetics and extent of embrittlement. The emphasis during the current year was on developing a procedure and correlations for predicting fracture toughness J-R curves of aged cast stainless steels from known material information. The present analysis has focused on developing correlations for the fracture properties in terms of material information that can be determined from the certified material test record (CMTR) and on ensuring that the correlations are adequately conservative for structurally weak materials

  18. Embrittlement phenomenon of Ag core MP35N cable as lead conductor in medical device.

    Wang, Ling; Li, Bernie; Zhang, Haitao

    2013-02-01

    Ag core MP35N (Ag/MP35N) wire has been used in lead electric conductor wires in the medical device industry for many years. Recently it was noticed that the combination of silver and MP35N restricts its wire drawing process. The annealing temperature in Ag/MP35N has to be lower than the melting temperature of pure Ag (960 °C), which cannot fully anneal MP35N. The lower annealing temperature results in a highly cold worked MP35N, which significantly reduces Ag/MP35N ductility. The embrittlement phenomenon of Ag/MP35N cable was observed in tension and bending deformation. The effect of the embrittlement on the wire flex fatigue life was evaluated using a newly developed flex fatigue testing method. The Ag/MP35N cable fatigue results was analyzed with a Coffin-Manson approach and compared to the MP35N cable fatigue results. The root causes of the Ag/Mp35N embrittlement phenomenon are discussed. Copyright © 2012 Elsevier Ltd. All rights reserved.

  19. The effect of gamma irradiation on bacteria in stored rice

    Kamaruzzaman Sijam.

    1987-01-01

    The effect of gamma irradiation on bacteria was studied for reducing the total microbial numbers that contaminating raw product under storage. Different storage packages of rice samples were irradiated at various levels of dosage. The results of bacterial isolation, total bacterial count and the isolation of bacterial food pathogenus were discussed. It was observed that the presence of bacteria colonies was suppressed by the presence of yeast and moulds eventhough the number of them decreased as the irradiation dosage levels were increased. (A.J.)

  20. UV-irradiation effects on polyester nuclear track detector

    Agarwal, Chhavi; Kalsi, P.C.

    2010-01-01

    The effects of UV irradiation (λ=254 nm) on polyester nuclear track detector have been investigated employing bulk-etch technique, UV-visible spectrophotometry and infra-red spectrometry (FTIR). The activation energy values for bulk-etching were found to decrease with the UV-irradiation time indicating the scission of the polymer. Not much shift in the absorption edge due to UV irradiation was seen in the UV-visible spectra. FTIR studies also indicate the scission of the chemical bonds, thereby further validating the bulk-etch rate results.

  1. Effect of gamma irradiation on quality of dried potato

    Wang, J.; Chao, Y.

    2003-01-01

    The objectives of this study were to obtain the effect of gamma irradiation on the quality of dried potato. Experiments were conducted to study the influence of different doses, air temperatures, slice thickness of potatoes on the dehydration rate, appearance quality (L-values), vitamin C content, and the rehydration ratio of dried potatoes. The greater the dose, the higher the dehydration rate, the lesser the vitamin C content, and the lower the rehydration ratio. The L-values for low-dose irradiation was greater than that for non-irradiated potatoes

  2. Effect of gamma irradiation on quality of dried potato

    Wang, J. E-mail: jwang@zju.edu.cn; Chao, Y

    2003-03-01

    The objectives of this study were to obtain the effect of gamma irradiation on the quality of dried potato. Experiments were conducted to study the influence of different doses, air temperatures, slice thickness of potatoes on the dehydration rate, appearance quality (L-values), vitamin C content, and the rehydration ratio of dried potatoes. The greater the dose, the higher the dehydration rate, the lesser the vitamin C content, and the lower the rehydration ratio. The L-values for low-dose irradiation was greater than that for non-irradiated potatoes.

  3. Effect of gamma irradiation on some nutritional factors of rice

    Mohamad Khan Ayob; Osman Hassan.

    1987-01-01

    The effect of gamma irradiation and types of packaging material used (namely: gunny sack, heavy duty polyethylene, woven laminated bags) on moisture content, gel viscosity and reducing sugar of rice was observed. Moisture content, gel viscosity and reducing sugar were determined by drying method, brookfield viscometer and Nelson method, respectively. The results showed that moisture and reducing sugar content were not significantly affected by types of material and irradiation doses. On the other hand gel viscosity was greatly influenced by irradiation doses and storage time. (A.J.)

  4. Effect of gamma irradiation on cefotaxime in the solid state

    Zegota, H.; Koprowski, M.; Zegota, A. [Technical Univ., Lodz (Poland). Inst. of Applied Radiation Chemistry

    1995-02-01

    The effect of {gamma}-irradiation on cefotaxime, a member of the third generation of cephalosporins, has been investigated by using different spectroscopic, chromatographic and microbiological analytical methods. Cefotaxime sodium salt was irradiated in dry state in the range of sterilization doses from 5.85 to 46.8 kGy. According to the results obtained, the degree of cefotaxime alterations was lower than 1%, even for the higher radiation dose used. Trace amounts of antibiotic radiolysis products have been found by HPLC. The microbiological assay carried out using E. coli test strain reveal that the activity of irradiated cefotaxime did not decrease. (author).

  5. Effect of total lymphoid irradiation in chronic progressive multiple sclerosis

    Cook, S.D.; Devereux, C.; Troiano, R.; Hafstein, M.P.; Zito, G.; Hernandez, E.; Lavenhar, M.; Vidaver, R.; Dowling, P.C.

    1986-01-01

    Total lymphoid irradiation (TLI; 1980 cGy) or sham irradiation was given to 40 patients with chronic progressive multiple sclerosis (MS) in a prospective, randomised, double-blind study. During mean follow-up of 21 months, MS patients treated with TLI has less functional decline than sham-irradiated MS patients (p<0.01). A significant relation was noted between absolute blood lymphocyte counts in the first year after TLI and subsequent course, patients with higher lymphocyte counts generally having a worse prognosis (p<0.01). TLI was well tolerated and associated with only mild short-term, and to date, long-term side-effects. (author)

  6. Neutron irradiation effect on thermomechanical properties of shape memory alloys

    Abramov, V.Ya.; Ionajtis, R.R.; Kotov, V.V.; Loguntsev, E.N.; Ushakov, V.P.

    1996-01-01

    Alloys of Ti-Ni, Ti-Ni-Pd, Fe-Mn-Si, Mn-Cu-Cr, Mn-Cu, Cu-Al-Mn, Cu-Al-Ni systems are investigated after irradiation in IVV-2M reactor at various temperatures with neutron fluence of 10 19 - 10 20 cm -2 . The degradation of shape memory effect in titanium nickelide base alloys is revealed after irradiation. Mn-Cu and Mn-Cu-Cr alloys show the best results. Trends in shape memory alloy behaviour depending on irradiation temperature are found. A consideration is given to the possibility of using these alloys for components of power reactor control and protection systems [ru

  7. The effect of ion irradiation on inert gas bubble mobility

    Alexander, D.E.; Birtcher, R.C.

    1991-09-01

    The effect of Al ion irradiation on the mobility of Xe gas bubbles in Al thin films was investigated. Transmission electron microscopy was used to determine bubble diffusivities in films irradiated and/or annealed at 673K, 723K and 773K. Irradiation increased bubble diffusivity by a factor of 2--9 over that due to thermal annealing alone. The Arrhenius behavior and dose rate dependence of bubble diffusivity are consistent with a radiation enhanced diffusion phenomenon affecting a volume diffusion mechanism of bubble transport. 9 refs., 3 figs., 2 tabs

  8. Multiscale Modeling of Hydrogen Embrittlement for Multiphase Material

    Al-Jabr, Khalid A.

    2014-05-01

    Hydrogen Embrittlement (HE) is a very common failure mechanism induced crack propagation in materials that are utilized in oil and gas industry structural components and equipment. Considering the prediction of HE behavior, which is suggested in this study, is one technique of monitoring HE of equipment in service. Therefore, multi-scale constitutive models that account for the failure in polycrystalline Body Centered Cubic (BCC) materials due to hydrogen embrittlement are developed. The polycrystalline material is modeled as two-phase materials consisting of a grain interior (GI) phase and a grain boundary (GB) phase. In the first part of this work, the hydrogen concentration in the GI (Cgi) and the GB (Cgb) as well as the hydrogen distribution in each phase, were calculated and modeled by using kinetic regime-A and C, respectively. In the second part of this work, this dissertation captures the adverse effects of hydrogen concentration, in each phase, in micro/meso and macro-scale models on the mechanical behavior of steel; e.g. tensile strength and critical porosity. The models predict the damage mechanisms and the reduction in the ultimate strength profile of a notched, round bar under tension for different hydrogen concentrations as observed in the experimental data available in the literature for steels. Moreover, the study outcomes are supported by the experimental data of the Fractography and HE indices investigation. In addition to the aforementioned continuum model, this work employs the Molecular Dynamics (MD) simulations to provide information regarding bond formulation and breaking. The MD analyses are conducted for both single grain and polycrystalline BCC iron with different amounts of hydrogen and different size of nano-voids. The simulations show that the hydrogen atoms could form the transmission in materials configuration from BCC to FCC (Face Centered Cubic) and HCP (Hexagonal Close Packed). They also suggest the preferred sites of hydrogen for

  9. Embrittlement and life prediction of aged duplex stainless steel

    Kuwano, Hisashi

    1996-01-01

    The stainless steel, for which the durability for long term in high temperature corrosive environment is demanded, is a complex plural alloy. Cr heightens the oxidation resistance, Ni improves the ductility and impact characteristics, Si improves the fluidity of the melted alloy and heightens the resistance to stress corrosion cracking, and Mo suppresses the pitting due to chlorine ions. These alloy elements are in the state of nonequilibrium solid solution in Fe base at practical temperature, and cause aging phenomena such as segregation, concentration abnormality and precipitation during the use for long term. The characteristics of stainless steel deteriorate due to this. Two-phase stainless cast steel, the example of the embrittlement of the material for an actual machine, the accelerated test of embrittlement, the activation energy for embrittlement, and as the mechanism of aging embrittlement, the spinodal decomposition of ferrite, the precipitation of G phase and the precipitation of carbides and nitrides are described. Also in the welded parts of austenitic stainless steel, delta-ferrite is formed during cooling, therefore, the condition is nearly same as two-phase stainless steel, and the embrittlement due to long term aging occurs. (K.I.)

  10. Irradiation effects in superconductor oxides. Effets d'irradiation dans les oxydes supraconducteurs

    Rullier-Albenque, F; Konczykowski, M [CEA-Ecole Polytechnique, 91 - Palaiseau (France). Lab. d' Etudes des Solides Irradies

    1993-01-01

    Several effects of irradiation on the 92 K - oxide superconductor YBa[sub 2]Cu[sub 3]O[sub 7] are reported. Whatever irradiation type, the critical temperature T[sub c] is found to decrease and the resistivity to increase. At sufficiently high damage levels, YBa[sub 2]Cu[sub 3]O[sub 7] is no longer superconducting and even displays a semiconducting-like behaviour. The alterations of superconducting properties are clearly related to oxygen defects - in the CuO[sub 2] planes or CuO chains... but we have shown experimentally that copper defects are also important. Magnetic properties of YBa[sub 2]Cu[sub 3]O[sub 7] in mixed state are also very sensitive to irradiation. By pinning the flux lines, irradiation defects can considerably increase the critical current density j[sub c]. At present, irradiations by highly energetic heavy ions (6 GeV Pb for instance), which produce cylindrical tubes of amorphous material (latent tracks) throughout the whole thickness of the samples, are probably the most efficient way to enhance j[sub c]. (Author). 18 refs., 7 figs.

  11. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Hennion, A

    1999-03-15

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  12. Immunological effects of irradiation: waiting for a model

    Doria, G.

    1979-01-01

    Decreased resistance to pathogens, is an effect of ionizing radiations, is largely mediated by impairment of the specific immune response. It has been established that the antibody response is depressed when the antigen is injected shortly before or immediately after irradiation. Recovery of the response in animals immunized after sublethal irradiation starts after about 1 week and may be complete 2 months after irradiation. If animals are immunized a few or several days before radiation exposure, some parameters of the antibody response are unaffected while others may display lower or higher values than normal, depending on the nature and physical form of the antigen. Recent studies have described the effects of whole-body irradiation on antibody ability. Since this antibody property affects the stability of immune complexes and, therefore, the antibody's ability to neutralize viruses and toxins, affinity was felt to be of importance to counteract infections in irradiated animals. Antibody affinity was found up to 20 times greater in irradiated than in control mice when antigen was injected 1 to 5 days before or 2 hr to 8 weeks after a sublethal dose of x-rays. Recovery profiles of mitotic responses of spleen cells to PHA, ConA, or LPS suggest that the enhancement of antibody affinity in irradiated mice could result from a relative lack of suppressor T cells. Dysfunctions of the immune system of irradiated animals can be attributed to alterations of the populations of immunologically competent cells. Unlike auxiliary cells, lymphocytes are extremely sensitive to ionizing radiations. Subpopulations of T and B lymphocytes appear to have different radiosensitivities and to develop with different efficiencies in irradiated animals. B cells seem to be more radiosensitive and to recover faster than T cells

  13. Research on sprout inhibition effect of refrigerated garlic by irradiation

    Zhang Xuan; He Jianzhong; Li Ruijun

    2005-01-01

    This paper researches the sprout inhibition effect by irradiation on refrigerated garlic. The results shows that, the garlic is still in the period of dormancy within 7 days after taken out from the refrigerated warehouse, and irradiation have a good sprout inhibition effect on it. The irradiation dose is 40-90 Gy, the same as that of the post harvest irradiation treatment on garlic. Refrigerate the Zhongmu Garlic (at -2 degree C-0 degree C) until the middle ten days of February the next year, place it at the room temperature (10 degree C-15 degree C) for 1-7 days after taking it out of the warehouse, then use 60 Co γ-ray to irradiate it until the absorbed dose reaches 40-90 Gy, the sprout inhibition effect can be realized. The test also indicates that the deposited time after taking out of the refrigerated warehouse is crucial to the sprout inhibition effect of refrigerated garlic by irradiation. (authors)

  14. Intestinal uptake of bile acids: effect of external abdominal irradiation

    Thomson, A.B.R.; Cheeseman, C.I.; Walker, K.

    1984-01-01

    Abdominal irradiation has recently been shown to influence the uptake of hexoses, amino acids, fatty acids and cholesterol into the jejunum of rats. The present studies were undertaken with a previously validated in vitro technique to determine the effect of abdominal irradiation from a cesium source on the rates of uptake of six bile acids into the jejunum, ileum, and colon. The results show that: 1) there likely are multiple ileal carriers for bile acids: 2) abdominal irradiation has a variable effect on these carriers; 3) the passive permeability to bile acids varies with the bile acid and with the site along the intestine; and 4) abdominal irradiation is associated with a rise in the colonic permeability to only some bile acids

  15. Effective suppression of bystander effects by DMSO treatment of irradiated CHO cells

    Kashino, Genro; Prise, K.M.; Suzuki, Keiji

    2007-01-01

    Evidence is accumulating that irradiated cells produce some signals which interact with non-exposed cells in the same population via a bystander effect. Here, we examined whether dimethyl sulfoxide (DMSO) is effective in suppressing radiation induced bystander effects in Chinese hamster ovary (CHO) and repair deficient xrs5 cells. When 1 Gy-irradiated CHO cells were treated with 0.5% DMSO for 1 hr before irradiation, the induction of micronuclei in irradiated cells was suppressed to 80% of that in non-treated irradiated cells. The suppressive effect of DMSO on the formation of bystander signals was examined and the results demonstrated that 0.5% DMSO treatment of irradiated cells completely suppressed the induction of micronuclei by the bystander effect in non-irradiated cells. It is suggested that irradiated cells ceased signal formation for bystander effects by the action of DMSO. To determine the involvement of reactive oxygen species on the formation of bystander signals, we examined oxidative stress levels using the 2',7'-dichlorofluorescein (DCFH) staining method in irradiated populations. The results showed that the treatment of irradiated cells with 0.5% DMSO did not suppress oxidative stress levels. These results suggest that the prevention of oxidative stress is independent of the suppressive effect of DMSO on the formation of the bystander signal in irradiated cells. It is suggested that increased reactive oxygen species (ROS) in irradiated cells is not a substantial trigger of a bystander signal. (author)

  16. Effect of neutron irradiation on select MAX phases

    Tallman, Darin J.; Hoffman, Elizabeth N.; Caspi, El’ad N.; Garcia-Diaz, Brenda L.; Kohse, Gordon; Sindelar, Robert L.; Barsoum, Michel W.

    2015-01-01

    Herein we report on the effect of neutron irradiation – of up to 0.1 displacements per atom at 360(20) °C or 695(25) °C – on polycrystalline samples of Ti 3 AlC 2 , Ti 2 AlC, Ti 3 SiC 2 and Ti 2 AlN. Rietveld refinement of X-ray diffraction patterns of the irradiated samples showed irradiation-enhanced dissociation into TiC of the Ti 3 AlC 2 and Ti 3 SiC 2 phases, most prominently in the former. Ti 2 AlN also showed an increase in TiN content, as well as Ti 4 AlN 3 after irradiation. In contrast, Ti 2 AlC was quite stable under these irradiation conditions. Dislocation loops are seen to form in Ti 2 AlC and Ti 3 AlC 2 after irradiation at 360(20) °C. The room temperature electrical resistivity of all samples increased by an order of magnitude after irradiation at 360(20) °C, but only by 25% after 695(25) °C, providing evidence for the MAX phases’ dynamic recovery at temperatures as low at 695(25) °C. Based on these preliminary results, it appears that Ti 2 AlC and Ti 3 SiC 2 are the more promising materials for high-temperature nuclear applications

  17. Effect of maize seed laser irradiation on plant photosynthetic activity

    Antonov, M.; Stanev, V.; Velichkov, D.; Tsonev, Ts.

    1986-01-01

    Investigations were made with the two hybrids, H-708 and P x -20. The seeds were irradiated by a helium-neon quantum generator (L'vov-1 Electronica) with output power of 24 MW and 632.8 nm wave length. Once and twice irradiated seeds were sown on the 2nd, 5th and 10th day post irradiation. Changes in leaf area, chlorophyll content in the leaves, photosynthetic rate and its dependence on temperature and light, transpiration, stomatal resistance to CO 2 and total dry matter of the overground plant part were traced. Seed irradiation with laser rays did not affect the chlorophyll content of the leaves. The photosynthetic rate did not depend on the cultivar characteristics of the crop. Single and repeated irradiation of the hybrid H-708 in most case enhanced photosynthetic rate, but a similar effect was not observed in P x -20. Transpiration and CO 2 stomatal resistance were not equally affected by radiation. Laser rays enhanced the ability of the photosynthetic apparatus of the entire plants to use more efficiently high light intensities. The leaf area and the total plant dry matter increased in case of sowing on the 2nd and 5th day and a single irradiation and in case of sowing on the 5th and 10th day and twice repeated irradiations

  18. Irradiation effects of 11 MeV protons on ferritic steels

    Hamaguchi, Yoshikazu; Kuwano, Hisashi; Misawa, Toshihei

    1985-01-01

    It is considered that ferritic/martensitic steels are the candidate of the first wall materials for future fusion reactors. The most serious problem in the candidate materials is the loss of ductility due to the elevation of ductile-brittle transition temperature by the high dpa irradiation of neutrons. 14 MeV neutrons produced by D-T reaction cause high dpa damage and also produce large quantity of helium and hydrogen atoms in first wall materials. Those gas atoms also play an important role in the embrittlement of steels. The main purpose of this work was to simulate the behavior of hydrogen produced by the transmutation in the mechanical properties of ferritic steels when they were irradiated with 11 MeV protons. The experimental procedure and the results of hardness, the broadening of x-ray diffraction lines, Moessbauer spectroscopy and small punch test are reported. High energy protons of 10 - 20 MeV are suitable to the simulation experiment of 14 MeV neutron radiation damage. But the production of the active nuclei emitting high energy gamma ray and having long life, Co-56, is the most serious problem. Another difficulty is the control of irradiation temperature. A small irradiation chamber must be developed. (Kako, I.)

  19. The effect of helium, radiation damage and irradiation temperature on the mechanical properties of beryllium

    Fabritsiev, S.A. [D.V. Efremov Scientific Research Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S.

    1998-01-01

    In this work different RF beryllium grades were irradiated in the BOR-60 reactor to a dose of {approx}5-10 dpa at irradiation temperatures 350, 420, 500, 800degC. Irradiation at temperatures of 350-400degC is shown to result in Be hardening due to the accumulation of radiation defect complexes. Hardening is accompanied with a sharp drop in plasticity at T{sub test} {<=} 300degC. A strong anisotropy in plasticity has been found at a mechanical testing temperature of 400degC and this parameter may be preferable when the samples are cut crosswise to the pressing direction. High-temperature irradiation (T{sub irr} = 780degC) gives rise to large helium pores over the grain boundaries and smaller pores in the grain body. Fracture is brittle and intercrystallite at T{sub test} {>=} 600degC. Helium embrittlement is accompanied as well with a drop in the Be strength properties. (author)

  20. Effects of irradiation on the gelation properties of muscle protein

    Lin Xianping; Yang Wenge

    2014-01-01

    Gel properties of muscle protein are the important functional characteristics in meat and its products. which determine the meat products' unique quality. such as texture. Juiciness. fat content and sensory characteristics As a novel food preservation technique, irradiation may lead to changes in the composition and structure of protein molecule. and impact the gel forming ability and gelation properties of muscle protein. Based on the introduction of gel forming mechanism of muscle protein, effects of irradiation on the water holding capacity, mechanical properties and structure of muscle protein gel were reviewed in detail. High-dose irradiation could weaken the water holding capacity of muscle protein and result in the loss of meat juice. With different irradiation conditions or raw materials, influences of irradiation on the texture and theological properties of muscle protein gels are varied, and effects on the structure of muscle protein and its gel are more complex. Finally, the research trend of irradiation effects on the gelation properties of muscle protein is put forward. (authors)

  1. Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

    Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Odette, George R. [Univ. of California, Santa Barbara, CA (United States); Almirall, Nathan [Univ. of California, Santa Barbara, CA (United States); Robertson, Janet [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Server, W. L. [ATI Consulting, Pinehurst, NC (United States); Yamamoto, T. [Univ. of California, Santa Barbara, CA (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States)

    2017-05-01

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

  2. Microstructural stability of spark-plasma-sintered Wf/W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation

    M. Avello de Lama; M. Balden; H. Greuner; T. Höschen; J. Matejicek; J.H. You

    2017-01-01

    Tungsten is considered as the most suitable material for the plasma-facing armour of future fusion reactors. However, in spite of many advantageous properties, pure tungsten has a major drawback, namely, brittleness at lower temperatures and embrittlement by neutron irradiation. Tungsten fibre-reinforced tungsten (Wf/W) composites are thought to be a promising candidate material for armour owing to the pseudo-toughness effect which is based on controlled cracking of coated interfaces. In this...

  3. Hydrogen embrittlement susceptibility of laser-hardened 4140 steel

    Tsay, L.W.; Lin, Z.W. [Nat. Taiwan Ocean Univ., Keelung (Taiwan). Inst. of Mater. Eng.; Shiue, R.K. [Institute of Materials Sciences and Engineering, National Dong Hwa University, Hualien, Taiwan (Taiwan); Chen, C. [Institute of Materials Sciences and Engineering, National Taiwan University, Taipei, Taiwan (Taiwan)

    2000-10-15

    Slow strain rate tensile (SSRT) tests were performed to investigate the susceptibility to hydrogen embrittlement of laser-hardened AISI 4140 specimens in air, gaseous hydrogen and saturated H{sub 2}S solution. Experimental results indicated that round bar specimens with two parallel hardened bands on opposite sides along the loading axis (i.e. the PH specimens), exhibited a huge reduction in tensile ductility for all test environments. While circular-hardened (CH) specimens with 1 mm hardened depth and 6 mm wide within the gauge length were resistant to gaseous hydrogen embrittlement. However, fully hardened CH specimens became susceptible to hydrogen embrittlement for testing in air at a lower strain rate. The strength of CH specimens increased with decreasing the depth of hardened zones in a saturated H{sub 2}S solution. The premature failure of hardened zones in a susceptible environment caused the formation of brittle intergranular fracture and the decrease in tensile ductility. (orig.)

  4. Effect of endotoxin preparations (LPS) with irradiation decreased toxicity on the immune response of normal and irradiated rats

    Elekes, E; Bertok, L [Orszagos Frederic Joliot-Curie Sugarbiologiai es Sugaregeszsegugyi Kutato Intezet, Budapest (Hungary)

    1979-03-01

    A comparison of the immunostimulating effect of parent and radiodetoxified with 50, 100, 150 and 200 kGy (5, 10, 15 and 20 Mrad) /sup 60/Co ..gamma..-rays endotoxin preparations in normal and irradiated rats is given. By increasing the dose of irradiation the immunostimulating effect decreased. The preparations detoxified even with the highest (200 kGy) dose is characterized by a pronounced adjuvant effect in irradiated animals.

  5. Preventive and Therapeutic Effects of Propolis in Gamma Irradiated Rats

    Hamza, R.G.; El-Shahat, A.N.

    2011-01-01

    Ionizing radiation is known to stimulate the generation of oxygen radicals which destabilize organic molecules resulting in a decrease of the system's antioxidant potential. Propolis (bee glue) is a complex mixture of natural substances that exhibits a broad spectrum of biological activities. As the possibility exists that it may exert a radio protections role, the present study aimed to examine the preventive and therapeutic effects of propolis on the gamma irradiation-induced changes in antioxidant status and certain biochemical parameters. HPLC chromatography for analysis of propolis showed that the number of identified phenols was 6 compounds (natural antioxidants). Male albino rats were exposed to 6 Gy of gamma radiation. The efficiency of propolis was evaluated when propolis was administered orally to rats at a dose of 200 mg/kg as follow: non-irradiated rats received orally propolis extract for 6 weeks (positive control) and rats received orally propolis extract for 3 weeks before or after gamma irradiation. The obtained results revealed that propolis given to rats before gamma irradiation protect the hazardous effects of gamma irradiation. In addition, administration of propolis to gamma irradiated rats caused significant enhancement in hepatic antioxidant enzymes (glutathion reductase; GR and catalase; CAT) and total antioxidant capacity associated with a remarkable decrease in the level of lipid peroxidation (TBARS). Also, it significantly reduced the changes induced by gamma irradiation in the serum levels of glucose and liver enzymes; aminotransferases (AST, ALT) and alkaline phosphatase (ALP). In addition, a significant improvement was observed in the serum levels of total cholesterol (TC), triglycerides (TG), low density lipoprotein- cholesterol (LDL-C) and high density lipoprotein-cholesterol (HDL-C). In conclusion, the positive results obtained in the gamma irradiated rats given propolis indicated that propolis could be considered as effective

  6. Multiscale modelling of hydrogen embrittlement in zirconium alloys

    Majevadia, Jassel; Wenman, Mark; Balint, Daniel; Sutton, Adrian [Imperial College London (United Kingdom); Nazarov, Roman [MPIE, Dusseldorf (Germany)

    2013-07-01

    Delayed Hydride Cracking (DHC) is a commonly occurring embrittlement phenomenon in zirconium alloy fuel cladding within Pressurized Water Reactors (PWRs). DHC is caused by the accumulation of hydrogen atoms taken up by the metal, and the formation of brittle hydrides in the vicinity of crack tips. The rate of crack growth is limited by the rate of hydrogen diffusion to the crack, which can be modelled by solving a stress driven diffusion equation that incorporates the elastic interaction between defects. This of interest in the present work. The elastic interaction is calculated by combining defect forces determined through Density Functional Theory (DFT) simulations, and an exact solution for the anisotropic elastic field of an edge dislocation in Zr. making it possible to determine the interaction energy without the need to simulate directly a hydrogen atom in the presence of a crack or dislocation, which is computationally prohibitive with DFT. The result of the elastic interaction energy calculations can be utilised to determine the segregation of hydrogen to a crack tip for varying crack tip geometries, and in the presence of other crystal defects. This is done by implementing a diffusion equation for hydrogen within a discrete dislocation dynamics simulation. In the present work a model has been developed to demonstrate the effect of a single dislocation on hydrogen diffusion to create a Cottrell atmosphere.

  7. The effect of helium generation and irradiation temperature on tritium release from neutron irradiated beryllium

    Kupriyanov, I.B.; Gorokhov, V.A.; Vlasov, V.V.; Kovalev, A.M.; Chakin, V.P.

    2004-01-01

    The effect of neutron irradiation condition on tritium release from beryllium is described in this paper. Beryllium samples were irradiated in the SM reactor with neutron fluence (E > 0.1 MeV) of (0.37-2.0) x 10 22 cm -2 at 70-100degC and 650-700degC. Mass-spectrometer technique was used in out of tritium release experiments during stepped-temperature anneal within a temperature range from 250 to 1300degC. The total amount of helium accumulated in irradiated beryllium samples varied from 521 appm to 3061 appm. The first signs of tritium release were detected at temperature of 406-553degC. It was shown that irradiation temperature and helium generation level significantly affect the tritium release. A fraction of 44 - 74 % of tritium content in samples irradiated at low temperature (70 - 100degC) is release from beryllium at an annealing temperature below 800degC, whereas for samples after high temperature irradiation (650 - 700 degC) tritium release did not exceed 14 %. Majority of tritium (∼68%) is released within a temperature range from 800 to 920 degC. The increase of helium generation from 521 appm to 3061 appm results in lowering the temperature of maximal tritium release rate and the upper temperature of tritium release from beryllium by 100-130degC and 200-240degC, correspondingly. On the basis of data obtained, the diffusion coefficients of tritium in beryllium were calculated. (author)

  8. Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

    Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

    2012-01-01

    Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different 10 B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

  9. Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

    Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-15

    Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

  10. Effects of bonding bakeout thermal cycles on pre- and post irradiation microstructures, physical, and mechanical properties of copper alloys

    Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J. [Pacific Northwest National Laboratory, Richland, WA (United States)

    1996-10-01

    At present, dispersion strengthened (DS) copper is being considered as the primary candidate material for the ITER first wall and divertor components. Recently, it was agreed among the ITER parties that a backup alloy should be selected from the two well known precipitation hardened copper alloys, CuCrZr and CuNiBe. It was therefore decided to carry out screening experiments to simulate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties, and electrical resistivity of CuCrZr and CuNiBe alloys. On the basis of the results of these experiments, one of the two alloys will be selected as a backup material. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime ageing, and bonding thermal cycle followed by reageing and the reactor bakeout treatment at 623K for 100 hours. Tensile specimens of the DS copper were also given the heat treatment corresponding to the bonding thermal cycle. A number of these heat treated specimens of CuCrZr, CuNiBe, and DS copper were neutron irradiated at 523K to a dose level of {approx}0.3 dpa (NRT) in the DR-3 reactor at Riso. Both unirradiated and irradiated specimens with the various heat treatments were tensile tested at 532K. The dislocation, precipitate and void microstructures and electrical resistivity of these specimens were also determined. Results of these investigations will be reported and discussed in terms of thermal and irradiation stability of precipitates and irradiation-induced precipitation and recovery of dislocation microstructure. Results show that the bonding and bakeout thermal cycles are not likely to have any serious deleterious effects on the performance of these alloys. The CuNiBe alloys were found to be susceptible to radiation-induced embrittlement, however, the exact mechanism is not yet known. It is thought that radiation-induced precipitation and segregation of the beryllium may be responsible.

  11. PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.; Taylor, B.J. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

  12. The decontamination effects of gamma irradiation on the edible gelatin

    Fu, Junjie; Shen, Weiqiao; Bao, Jinsong; Chen, Qinglong

    2000-01-01

    The decontamination effects of gamma irradiation on the edible gelatin were studied. The results indicated that the bacterium and mold in the gelatin decreased significantly with the dose of 5 kGy treatment. However, the content of crude protein, microelement, amino acid in the gelatin remained unchanged under the irradiation of 4 and 8 kGy. The viscosity of the gelatin decreased with the increase of the irradiation dose, but the gelatin with a dose of 5 kGy treatment still accorded with the standard of the second-order class. These results suggested that the optimum irradiation dose for edible gelatin for the purpose of decontamination was in the range 3-5 kGy. (author)

  13. Effect of irradiation on lysosomal enzyme activation in cultured macrophages

    Clarke, C.; Wills, E.D.

    1980-01-01

    The effect of γrays on lysosomal enzyme activity of normal and immune macrophages of DBA/2 mice cultured in vitro has been studied. A dose of 500 rad did not significantly affect lysosomal enzyme activity 3 hours after irradiation but caused the activity to increase to 1.4 times the control value 22.5 hours after irradiation. 22.5 hours after a dose of 3000 rad the enzyme activity increased to 2.5 times the control. Lysosomal enzyme activity of the macrophages was also markedly increased by immunization of the mice with D lymphoma cells, before culture in vitro, but irradiation of these cells with a dose of 500 rad caused a further increase in lysosomal enzyme activity. The results indicate that immunization and irradiation both cause stimulation of lysosomal enzyme activity in macrophages but that the mechanisms of activation are unlikely to be identical. (author)

  14. The effect of irradiation on the DNA of cauliflower

    Harmey, M.A.

    1991-01-01

    The cellular DNA is one of the components most affected by ionizing radiation. Lesions caused range from single and double stranded breaks to chemical modification of bases depending on the radiation dosage and the metabolic status of the tissue. In attempting to assess the DNA damage induced by irradiation of vegetables in a speedy and convenient manner, we examined the effect on the DNA by subjecting cauliflower to a dose of 1 kGy. If DNA is nicked by irradiation, the extent of the damage can be assessed by using DNA polymerase to repair the nicks. Comparisons were made between irradiated and non irradiated cauliflower and incorporation of 32 p deoxy GTP in the presence of the Klenow fragment of DNA polymerase measured

  15. Effect of irradiation decontamination on the qualities of green tea

    Zhu Jiating; Liu Chunquan; Yu Gang; Zhao Yongfu; Ji Ping; Jin Jie; Gu Guiqiang

    2005-01-01

    The purpose of this study was to analyze the effects of irradiation on the main chemical components, such as heavy metal elements, pesticide residues as well as sensory qualities of green tea. The results indicated that irradiation had no significant impact on proteins, tea polyphenols, theine and heavy metal elements, slight differences in the contents of soluble sugar and amino acids. The content of cypermethrin reduced with the increase of irradiation dose. The color, liquor color, flavor and aroma of the tea decoction changed slightly when irradiated at the dose lower than 5 kGy. It was concluded that the optimal doses for the purpose of green team decontamination was at the range of 3-5 kGy according to the analysis of various quality factors. (authors)

  16. Acute effects of gamma irradiation on vascular arterial tone

    Bourlier, V.; Diserbo, M.; Multon, E.; Verdetti, J.; Fatome, M.

    1995-01-01

    In rat aortic rings, we showed an increase in arterial tone during irradiation. This effect is acute reversible. This effect is only observed on pre-contracted rings and needs the integrity of vascular endothelium. The molecular mechanism of this effect is discussed. (author)

  17. Effects of Irradiation on Albite's Chemical Durability.

    Hsiao, Yi-Hsuan; La Plante, Erika Callagon; Krishnan, N M Anoop; Le Pape, Yann; Neithalath, Narayanan; Bauchy, Mathieu; Sant, Gaurav

    2017-10-19

    Albite (NaAlSi 3 O 8 ), a framework silicate of the plagioclase feldspar family and a common constituent of felsic rocks, is often present in the siliceous mineral aggregates that compose concrete. When exposed to radiation (e.g., in the form of neutrons) in nuclear power plants, the crystal structure of albite can undergo significant alterations. These alterations may degrade its chemical durability. Indeed, careful examinations of Ar + -implanted albite carried out using Fourier transform infrared spectroscopy (FTIR) and molecular dynamics simulations show that albite's crystal structure, upon irradiation, undergoes progressive disordering, resulting in an expansion in its molar volume (i.e., a reduction of density) and a reduction in the connectivity of its atomic network. This loss of network connectivity (i.e., rigidity) results in an enhancement of the aqueous dissolution rate of albite-measured using vertical scanning interferometry (VSI) in alkaline environments-by a factor of 20. This enhancement in the dissolution rate (i.e., reduction in chemical durability) of albite following irradiation has significant impacts on the durability of felsic rocks and of concrete containing them upon their exposure to radiation in nuclear power plant (NPP) environments.

  18. Physical properties of beryllium oxide - Irradiation effects

    Elston, J.; Caillat, R.

    1958-01-01

    This work has been carried out in view of determining several physical properties of hot-pressed beryllium oxide under various conditions and the change of these properties after irradiation. Special attention has been paid on to the measurement of the thermal conductivity coefficient and thermal diffusivity coefficient. Several designs for the measurement of the thermal conductivity coefficient have been achieved. They permit its determination between 50 and 300 deg. C, between 400 and 800 deg. C. Some measurements have been made above 1000 deg. C. In order to measure the thermal diffusivity coefficient, we heat a perfectly flat surface of a sample in such a way that the heat flux is modulated (amplitude and frequency being adjustable). The thermal diffusivity coefficient is deduced from the variations of temperature observed on several spots. Tensile strength; compressive strength; expansion coefficient; sound velocity and crystal parameters have been also measured. Some of the measurements have been carried out after neutron irradiation. Some data have been obtained on the change of the properties of beryllium oxide depending on the integrated neutron flux. (author) [fr

  19. Characterisation of irradiation effect on geo-polymers

    Chupin, Frederic

    2015-01-01

    This study aims to improve knowledge about the radiation effect on geo-polymer behavior in terms of dihydrogen release and general strength in order to consider them as an alternative to usual nuclear waste cementitious coating matrices. Using various characterization techniques (nitrogen adsorption, low temperature DSC, FTIR and 1 H NMR spectroscopy) and by means of simulation irradiations (gamma, heavy ions), it has been shown that all the water present in the geo-polymer could be radiolyzed and that there was a confinement effect on the water radiolysis under low LET irradiation, probably due to efficient energy transfers from the solid matrix to the interstitial solution. Three dihydrogen production rates have been identified with the absorbed dose, depending on the concentration of dissolved dioxygen and the dihydrogen accumulation in the geo-polymer matrix. The good mechanical strength of the geo-polymer has been shown up to 9 MGy under gamma irradiation and is due to its high stability under irradiation. This could be explained by the fast recombination of the defects observed by EPR spectroscopy. However, phase crystallization was revealed during irradiation with heavy ions, which may induce some weakening of the geo-polymer network under alpha irradiation. The overall results helped to understand the phenomenology in a waste package under storage conditions. (author) [fr

  20. Effects of irradiation on the components of implantable pacemakers

    Kawamura, S; Kuga, N; Shiba, T; Hirose, T; Fujimoto, H; Toyoshima, T; Hyodo, K; Matoba, M

    2003-01-01

    The purpose of this study was to examine the effects of irradiation on implantable pacemaker components. The pacemaker was divided into three components: lead wire and electrode, battery, and electrical circuit, and each component was irradiated by X-ray and electron beams, respectively. The pacemaker parameters were measured by both telemetry data of the programmer and directly measured data from the output terminal. The following results were obtained. For the lead wire and electrode, there was no effect on the pacemaker function due to irradiation by X-ray and electron beams. In the case of battery irradiation, there was no change in battery voltage or current up to 236 Gy X-ray dose. In the electrical circuit, the pacemaker reverted to the regular beating rate (fixed-rate mode) immediately after the start of X-ray irradiation, and it continued in this mode during irradiation. In patients with their own heartbeat rhythm, changing to the fixed-rate mode may cause dangerous conditions such as ventricular fib...