Search

The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory is an accelerator-based pulsed neutron source that produces high-energy spallation neutrons by bombarding liquid mercury flowing through a stainless steel target vessel. During operation the proton beam and spallation neutrons produce radiation damage in the AISI 316L austenitic stainless steel target vessel and water-cooled shroud. The beam pulses also cause rapid heating of the liquid mercury, which may produce cavitation erosion damage on the inner surface of the target vessel. The cavitation erosion rate is thought to be highly sensitive to beam power and predicted to be the primary life-limiting factor of the target module. Though cavitation erosion and radiation damage to the target vessel are expected to dictate its lifetime, the effects of radiation damage and cavitation erosion to target vessels in liquid metal spallation systems are not well known. Therefore preparations are being undertaken to perform post-irradiation examination (PIE) of the liquid mercury target vessel and water-cooled shroud after end-of-life occurs. An overview of the planned PIE for the SNS target vessel is presented here, including proposed techniques for specimen acquisition and subsequent material properties characterization.

A number of Ti-based alloys in both the mill-annealed and 20% cold-worked conditions were subjected to sonication conditions in Hg using a vibratory horn to assess relative cavitation–erosion resistance. Weight loss as a function of exposure time decreased monotonically with increasing hardness for all alloys/conditions examined, with Ti–6Al–4V (Grade 5) and Ti–6Al–2Sn–4Zr–2Mo yielding the best resistance to cavitation–erosion as evidenced by low weight losses and little or no tendency to form pits on the exposed surface. Unalloyed Ti (Grade 4) and Ti–0.12Pd (Grade 7) exhibited greater weight losses by a factor of about two and about five, respectively, with Ti–0.12Pd particularly prone to pitting development. The mean erosion rates of the best two Ti-alloys examined were about a factor of three higher than identically tested 316LN stainless steel following a low temperature carburizing treatment, but this difference is considered minor given that the rate for both materials is very low/manageable and represents a through-thickness property for the Ti-alloys. A nitriding surface treatment was also evaluated as a potential method to further increase the cavitation–erosion resistance of these alloys in Hg, but the selected treatment proved largely ineffective as measured by rapid weight loss. Recommendations for further work to evaluate the efficacy of Ti-based alloys for use in high-powered targets for the Spallation Neutron Source are given.

Because of their superior high temperature strength and corrosion properties, a set of Ni-base alloys has been proposed for various in-core applications in Gen IV reactor systems. However, irradiation-performance data for these alloys is either limited or non-existent. A review is presented of the irradiation-performance of a group of Ni-base alloys based upon data from fast breeder reactor programs conducted in the 1975–1985 timeframe with emphasis on the mechanisms involved in the loss of high temperature ductility and the breakdown in swelling resistance with increasing neutron dose. The implications of these data for the performance of the Gen IV Ni-base alloys are discussed and possible pathways to mitigate the effects of irradiation on alloy performance are outlined. A radical approach to designing radiation damage-resistant Ni alloys based upon recent advances in mechanical alloying is also described.

Room temperature cavitation tests of vacuum annealed type 316LN stainless steel were performed in pure mercury and in mercury with various amounts of metallic solute to evaluate potential mitigation of erosion/wastage. Tests were performed using an ultrasonic vibratory horn with specimens attached at the tip. All of the solutes examined, which included 5 wt% In, 10 wt% In, 4.4 wt% Cd, 2 wt% Ga, and a mixture that included 1 wt% each of Pb, Sn, and Zn, were found to increase cavitation-erosion as measured by increased weight loss and/or surface profile development compared to exposures for the same conditions in pure mercury. Qualitatively, each solute appeared to increase the tenacity of the post-test wetting of the Hg solutions and render the Hg mixture susceptible to manipulation of droplet shape.

An overview of the design and construction of the Spallation Neutron Source (SNS) is presented. Key facility performance parameters are summarized and plans for initial operation are described. Early efforts produced a conceptual design in 1997; the project itself was initiated in 1999, with the official groundbreaking taking place in December of 1999. As of April 2005 building construction was complete and the overall project was more than 90% complete. The design of the target and surrounds are finished and the first target was installed in June 2005. First beam on target is expected in June, 2006. The engineering design of the target region is described. The key systems comprise the mercury target, moderator and reflector assemblies, remote handling systems, utilities and shielding. Through interactions with the 1 GeV proton beam, the target, moderators and reflectors produce short pulse neutrons in thermal energy ranges, which are transported to a variety of neutron scattering instruments. The mercury target module itself is described in more detail. Materials issues are expected to govern the overall lifetime and have influenced the design, fabrication and planned operation. A wide range of materials research and development has been carried out to provide experimental data and analyses to ensure the satisfactory performance of the target and to set initial design conditions. Materials R&D concentrated mainly on cavitation erosion, radiation effects, and mercury compatibility issues, including investigations of the mechanical properties during exposure to mercury. Questions that would require future materials research are discussed.

An extensive fatigue testing program on 316LN stainless steel was recently carried out to support the design of the mercury target container for the spallation neutron source (SNS) that is currently under construction at the Oak Ridge National Laboratory in the United States. The major objective was to determine the effects of mercury on fatigue behavior. The S – N fatigue behavior of 316LN stainless steel is characterized by a family of bilinear fatigue curves which are dependent on frequency, environment, mean stress and cold work. Generally, fatigue life increases with decreasing stress and levels off in the high cycle region to an endurance limit below which the material will not fail. For fully reversed loading as well as tensile mean stress loading conditions mercury had no effect on endurance limit. However, at higher stresses a synergistic relationship between mercury and cyclic loading frequency was observed at low frequencies. As expected, fatigue life decreased with decreasing frequency, but the response was more pronounced in mercury compared with air. As a result of liquid metal embrittlement (LME), fracture surfaces of specimens tested in mercury showed widespread brittle intergranular cracking as opposed to typical transgranular cracking for specimens tested in air. For fully reversed loading (zero mean stress) the effect of mercury disappeared as frequency increased to 10 Hz. For mean stress conditions with R -ratios of 0.1 and 0.3, LME was still evident at 10 Hz, but at 700 Hz the effect of mercury had disappeared ( R = 0.1). Further, for higher R -ratios (0.5 and 0.75) fatigue curves for 10 Hz showed no environmental effect. Finally, cold working (20%) increased tensile strength and hardness, and improved fatigue resistance. Fatigue behavior at 10 and 700 Hz was similar and no environmental effect was observed.

Fusion reactors, advanced fission reactors and high power accelerator spallation targets subject materials to damaging particle irradiation. Although these technologies derive their utility from different nuclear reactions and divergent applications, they experience many common features. Further, the physical mechanisms of radiation response are cross-cutting. For example, swelling, phase instability, hardening, flow localization, and embrittlement must be understood in order to estimate component lifetimes. Additional commonalities include reliance on the same classes of materials and sometimes on the identical alloy for critical components. In addition, databases supporting designs are mainly derived from the same relatively few irradiation facilities and from similar types of experiments. Opportunities are examined for coordinated efforts. Emphasis is placed on the development of fundamental knowledge to support alloy design strategies for resistance to irradiation and to form a scientific basis to develop better materials.

The high-cycle fatigue behavior of Type 316 stainless steel (SS), the prime candidate target-container material for the spallation neutron source (SNS), was investigated in air and mercury at frequencies of 0.2 and 10 Hz with a R ratio of −1, and at 10 and 700 Hz with a R ratio of 0.1. Here R equals the ratio of the applied minimum to maximum loads during fatigue experiments. A decrease in the fatigue life in mercury was observed, relative to that in air, at 0.2 Hz. Correspondingly, intergranular fracture was found on the fracture surfaces of specimens tested in mercury at 0.2 Hz, which is a typical fracture mode caused by liquid metal embrittlement (LME). Heating by mechanical working was observed during fatigue tests at 10 Hz and a R of −1, and at 700 Hz and a R of 0.1, which resulted in great increases in specimen temperatures and shorter fatigue lives for large stress amplitudes (⩾210 MPa), relative to those in mercury. However, in the fatigue tests at 10 and 700 Hz, the fatigue lives in air with cooling and those in mercury seemed to be comparable, indicating little influence of the mercury. Thus, both specimen self-heating and LME need to be considered in understanding fatigue behavior of Type 316 SS in air and mercury.

Design of the mercury target system components for the spallation neutron source (SNS) requires data on high- and low-cycle fatigue behavior. The research and development program in progress includes determining the effects of mercury on the fatigue behavior of type 316 LN stainless steel, the primary material of choice for the target vessel. Uniaxial, load-controlled, fully reversed tension–compression R=−1 (minimum stress/maximum stress) fatigue tests have been conducted in air and mercury at room temperature employing constant amplitude sinusoidal loading at frequencies from 0.2 to 10 Hz. Stress amplitude versus fatigue life data (S–N curves) for both air and mercury show a sharp knee at approximately 1 million cycles indicating a fatigue endurance limit in either air or mercury around 240 MPa. Tensile mean stress (R=0.1) lowers the endurance limit to 160 MPa. Lower frequency and mercury environment had some impact (degradation) on fatigue life of type 316 LN stainless steel at high stress levels (i.e., stresses considerably above the apparent fatigue limit). Test results for high mean stress conditions (R=0.3, 0.5, and 0.75) at a cyclic frequency of 10 Hz exhibited further reductions in the endurance limit.

In the Spallation Neutron Source target, the structural material will be exposed to intense pulsed fluxes of high-energy protons and neutrons, which produce radiation damage. These pulsed fluxes also lead to pressure pulses created by beam heating. In turn, the pressure pulses give rise to fluctuating stresses in the 316 LN austenitic stainless steel target vessel, and to cavitation in the liquid mercury spallation target. Corrosion reactions and related changes in mechanical properties also may occur through contact with flowing mercury. We describe the materials research and development program for the spallation target. The program covers the areas of cavitation erosion, radiation effects, and compatibility. Cavitation erosion work includes pressure wave tests at the LANSCE proton accelerator, as well as laboratory tests that simulate aspects of the actual in-beam exposures. Materials irradiations are being carried out in spallation environments at high-energy and high-power proton accelerators. Other experiments are conducted at irradiation facilities that simulate aspects of spallation conditions. Extensive radiation damage and transmutation calculations supplement these experiments. Compatibility work includes both thermal convection and pumped flow loop tests to examine temperature gradient mass transfer, as well as fatigue and tensile tests in contact with Hg. Based on the information developed for radiation effects and compatibility with mercury, our analysis indicates that the target will meet its intended service requirements. In the past year and one half the new issue of cavitation erosion has been included in the program. Both in-beam and laboratory experiments indicate that cavitation erosion may occur in the target. The highest priority activity is now to determine whether cavitation erosion will limit target lifetime to a level below the lifetime limit set by radiation effects.

As an assessment for a possible accelerator beam line window material for the US Spallation Neutron Source (SNS) target, performance, radiation-induced hardening and microstructural evolution in Inconel 718 were investigated in both solution annealed (SA) and precipitation hardened (PH) conditions. Irradiations were carried out using 3.5 MeV Fe+, 370 keV He+ and 180 keV H+ either singly or simultaneously at 200 °C to simulate the damage and He/H production in the SNS target vessel wall. This resulted in systematic hardening in SA Inconel and gradual net softening in the PH material. TEM microstructural analysis showed the hardening was associated with the formation of small loop and faulted loop structures. Helium-irradiated specimens included more loops and cavities than Fe+ ion-irradiated specimens. Softening of the PH material was due to dissolution of the γ′/γ″ precipitates. High doses of helium were implanted in order to study the effect of high retention of gaseous transmutation products. Simultaneous with the hardening and/or softening due to the displacement damage cascade, helium filled cavities produced additional hardening at high concentrations.

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr–2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54–2.53 dpa at 30–100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr–2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress–strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13–18% and the ductility by 8–36%. The effect of test temperature for the 9Cr–2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress–strain data predicted positive strain hardening during plastic instability.

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

The effects of displacement damage and trapped helium on deformation microstructures in AISI 316 LN austenitic stainless steel were studied by applying a newly developed disk bend method to specimens irradiated with 360 keV He ions at 200°C. Radiation damage microstructures consisted of an intimate mix of black dots, dislocation loops, and very small helium filled cavities. In the unirradiated specimens, the deformation mode upon straining was planar glide with cross-slip. With increasing dose, cross-slip was progressively restricted. Correspondingly, deformation microstructure changed from dislocation network dominant to channeling dominant. The channel bands were composed of piled-up dislocations, stacking faults, and twinned layers.

In this study, the focus is on the possible effect of mercury and transmuted atom recoils that are implanted into the wall of the stainless steel container vessel at SNS. Two computer codes were used: Los Alamos high energy transport (LAHET) and the stopping and range of ions in matter (SRIM). LAHET provided information of the energy distribution of the recoils upon bombardment. Also, SRIM simulated the transport of recoils through the mercury layer adjacent to the stainless steel wall and into the wall itself, and provided recoil and displacement concentration profiles in the wall. The calculated recoil and displacement concentrations in the near-surface region of the wall appear to be quite significant. The recoil and displacement concentrations per year at the surface are determined to be about 0.0015 recoil atoms/wall atom and about 17 dpa at the center of an incident proton beam. These concentrations fall to about one-half of the surface values at a penetration distance of about 0.1 μm, and the concentrations become negligibly small at 0.5 μm and beyond.

An overview of the research and development program for the Spallation Neutron Source (SNS) is presented. The materials-related efforts in target development are emphasized in order to provide a perspective for a number of specialized papers that are included in these proceedings. We give a brief introduction and historical sketch of the SNS project. Part of the materials R&D consists of calculations of radiation damage and of transmutation rates. He and H are considered to be the most important transmutation products. Radiation effects and Hg compatibility investigations make up the major part of the experimental effort. In the former, spallation irradiations are carried out in the LANSCE at Los Alamos National Laboratory and in the SINQ at the Paul Scherrer Institute. Irradiations that simulate aspects of a spallation environment are included to extend the parameter space of the spallation irradiations. The simulations are carried out at the low energy (MeV) accelerators of the TIF facility and at the HFIR reactor, both located at Oak Ridge National Laboratory. Irradiated specimens are tested for changes in mechanical properties and are characterized with respect to microstructural changes by transmission electron microscopy. The compatibility experiments cover both the effects of Hg on behavior in mechanical properties tests, and the effects of flowing Hg on mass transfer in target structural materials. The results of this extensive program of materials work indicate that the target design and materials performance will meet their intended service.

In the past, our calculations of radiation damage (concentrations of displacements, helium atoms, and hydrogen atoms) to the 316 stainless steel (316SS) container vessel at the spallation neutron source (SNS) mercury target dealt with the average damage rates in a volume at the nose of the vessel. This paper describes an attempt to improve the accuracy of estimates of the damage rates at the center of the proton beam where the damage rates are expected to be the highest. Four series of calculations (Series I–IV) were conducted to determine the damage rates within volumes (tally volumes) that varied systematically in location in the vessel. This permitted extrapolation to the rates at the tip of the vessel nose ( Z =0) and at the center of the proton beam ( X = Y =0). The total damage rates due to protons and neutrons were found to be: 36 dpa/yr, 1400 appmHe/yr, and 20 000 appmH/yr. In addition, insight was gained into how the damage rates vary with position in the vessel nose and at locations further downstream in the vessel.

The Fourth International Workshop on Spallation Materials Technology (IWSMT-4) took place on 8–13 October in Schruns, Austria. The present volume contains the proceedings of this workshop. 53 scientists from Europe, USA and Japan presented 44 papers covering a wide range of topics including the status of spallation target R&D activities, effects of radiation damage, hydrogen, helium and other transmutation impurities in materials, particle transport calculations, heavy liquid metal corrosion and compatibility, and materials engineering. The main progress, with an emphasis on materials technology in different projects, namely the US Spallation Neutron Source (SNS), the European Spallation Source (ESS), the Japanese Spallation Neutron Source (JSNS), the Swiss Spallation Neutron Source (SINQ) and the US Accelerator Production of Tritium facility (APT) was overviewed. Progress from the European TECLA and SPIRE programs was described. The achievements of investigations in the last two years on mechanical properties and microstructures of materials under high-energy proton (spallation), neutron and low energy ion irradiations were reported. The latest results from studies on corrosion and embrittlement effects of heavy liquid metals (Hg, Pb–Bi and Pb) were presented. Some critical issues of liquid metal technology related to accelerator driven systems (ADS), e.g. oxygen control and oxygen measurements were discussed.

Inconel 718 is a material under consideration for areas in the target region of the spallation neutron source (SNS), now under construction at Oak Ridge National Laboratory (ORNL) in the US. In these positions, displacement damage from protons and neutrons will affect the mechanical properties. In addition, significant amounts of helium and hydrogen will build up in the material due to transmutation reactions. Nanoindentation measurements of solution-annealed (SA) Inconel 718 specimens, implanted with Fe-, He-, and H-ions to simulate SNS target radiation conditions, have shown that hardening occurs due to ion-induced displacement damage as well as due to the build-up of helium bubbles in the irradiated layer. Precipitation-hardened (PH) Inconel 718 also exhibited hardening by helium build-up but showed softening as a function of displacement damage due to dissolution of the γ′ and γ″ precipitates.

A disk bend test technique has been developed to study deformation mechanisms as well as mechanical properties. In the disk bend test, a transmission electron microscopy (TEM) disk size specimen of 3 mm diameter ×0.25 mm thick is clamped around its rim in a circular holder and indented with a tungsten carbide ball of 1 mm diameter on its back face. AISI 316LN austenitic stainless steel and 9Cr–2WVTa ferritic/martensitic steel were selected as test materials. A model was developed to determine the average plastic strain and surface plastic strain in the disk bend test. The deformation regimes of the plastic strain versus deflection curves corresponded to those of the load versus deflection curves. The stress state of the disk bend deformation was analyzed for the two test materials and compared with those of other mechanical tests such as uniaxial tensile, compact tension, and ball indentation tests. Slip line features at the deformed surface and the corresponding TEM microstructures were examined for both tensile and disk bend specimens. Differences and similarities in deformation between the disk bend and the tensile tests are described.

In the construction materials surrounding the spallation neutron source (SNS) mercury target, considerable quantities of transmutation products, particularly hydrogen and helium, will be generated due to the exposure to a high flux of 1 GeV protons and associated neutrons. In an effort to investigate the effects of high helium, therefore, bubble formation and defect clustering processes in AISI 316 LN austenitic steel were studied as a function of helium concentration and displacement damage dose with 360 keV He + and 3500 keV Fe + ion beams at 200°C. Helium irradiation was less effective in producing defects such as black dots and dislocation loops than Fe + ion irradiation at equivalent displacement dose. On the other hand, the formation of helium bubbles produced a strong depressive effect on the growth of loops and the evolution of line dislocations. The results indicated that the effect of helium bubbles was augmented as the bubble number density and size increased with increasing helium beyond 1 atomic percent (at.%). In such a case, the effect of helium bubbles can be more important than that of radiation-induced defects on the evolution of microstructure and the change in mechanical properties.

For an assessment of the future US spallation neutron source (SNS) target performance, radiation induced hardening and microstructural evolution were investigated as a function of ion dose for EC316LN stainless steel. Irradiation was carried out using 3.5 MeV Fe+, 360 keV He+, and 180 keV H+ simultaneous ion-beams at 200°C to simulate the damage, He and H production in the SNS target vessel wall. At low dose ( 20% elongation) was maintained up to a dose level of ≃10 dpa .

The precipitate phase stability in Fe–15Ni–13Cr base austenitic alloys was investigated as a function of minor alloying additions after thermally aging at 600°C and 675°C for times ranging from 24 h to one year. Seven major precipitate phases were found in aged specimens, including M23C6, Laves, Eta (η), TiO, NbC, MC, and M2P. The types and amounts of precipitate phases varied with alloying element additions, aging temperature, and aging time. By analyzing the composition of each individual particle, it was possible to determine the essential constituent elements for each phase. From this information, a strategy to promote or suppress certain precipitate phases was developed. Among the seven phases, the most desirable precipitate phases were considered to be MC and M2P, because these particles form on a fine scale with a high number density and, therefore, can serve as effective gas atom trap sites under irradiation.

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

In an effort to develop alloys for fission and fusion reactor applications, 28Fe–15Ni–13Cr base alloys were fabricated by adding various combinations of the minor alloying elements, Mo, Ti, C, Si, P, Nb, and B. The results showed that a significant fraction of undesirable residual oxygen was removed as oxides when Ti, C, and Si were added. Accordingly, the concentrations of the latter three essential alloying elements were reduced also. Among these elements, Ti was the strongest oxide former, but the largest oxygen removal (over 80%) was observed when carbon was added alone without Ti, since gaseous CO boiled off during melting. This paper recommends an alloy melting procedure to mitigate solute losses while reducing the undesirable residual oxygen. In this work, 14 different types of precipitate phases were identified. Compositions of precipitate phases and their crystallographic data are documented. Finally, stability of precipitate phases was examined in view of Gibbs free energy of formation.

To simulate radiation damage under a future Spallation Neutron Source (SNS) environment, irradiation experiments were conducted on a candidate 9Cr–2WVTa ferritic/martensitic steel using the Triple Ion Facility (TIF) at ORNL. Irradiation was conducted in single, dual and triple ion beam modes using 3.5 MeV Fe ++ , 360 keV He + , and 180 keV H + at 80°C, 200°C and 350°C. These irradiations produced various defects comprising black dots, dislocation loops, line dislocations, and gas bubbles, which led to hardening. The largest increase in hardness, over 63%, was observed after 50 dpa for triple beam irradiation conditions, revealing that both He and H are augmenting the hardening. Hardness increased less than 30% after 30 dpa at 200°C by triple beams, compatible with neutron irradiation data from previous work which showed about a 30% increase in yield strength after 27.2 dpa at 365°C. However, the very large concentrations of gas bubbles in the matrix and on lath and grain boundaries after these simulated SNS irradiations make predictions of fracture behavior from fission reactor irradiations to spallation target conditions inadvisable.

Fast neutrons are usually considered the source of various radiation effects caused by atom displacements, such as embrittlement, swelling, and creep. However, other reactions may contribute to atom displacements under certain conditions. These additional sources of displacements include those caused by thermal neutron capture recoils, γ-induced energetic electrons, and energetic particles emerging from transmutation reactions. In reactor vessels, for example, special circumstances may be encountered where these reactions become significant or even dominant with respect to fast-neutron-induced displacements. Key considerations describing these relative contributions are described. Inequalities are derived giving requirements among materials parameters and irradiation conditions to make each process significant with respect to fast neutrons. An example of the application of these conditions is given, covering the explanation of ‘early’ embrittlement in the HFIR reactor vessel.

When a material is irradiated with energetic neutrons or charged particles, a complex sequence of reactions takes place. Structure, composition, and properties are altered over an extremely wide scale, spanning atomic defects, meso-scale microstructures and macroscopic properties. Particularly interesting are radiation-induced dimensional changes. Such changes, which can occur in engineered components of fission and fusion reactors on a scale of meters, are driven by lattice defects at the subnanometer level. Because of their technological importance and their high scientific challenge, the dimensional changes termed radiation-induced swelling and creep have elicited sustained intensive research by basic and applied materials scientists for many years. The present paper is intended as a brief tutorial on salient features of this work. The presentation is divided into three parts. Background is first sketched emphasizing experimentally observed features and applications. Next, the theoretical framework and specific models that have been developed to understand radiation-induced swelling and creep in isotropic materials are described. Lastly, selected experiments designed and/or interpreted in terms of theory are highlighted to illustrate the current state of understanding of the physical bases of these phenomena.

Investigations underway at Oak Ridge National Laboratory (ORNL) into reasons for the accelerated embrittlement of surveillance specimens of ferritic steels irradiated at 50°C at the High Flux Isotope Reactor (HFIR) pressure vessel are described. Originally, the major suspects for the precocious embrittlement were a highly thermalized neutron spectrum, a low displacement rate, and the impurities boron and copper. Each of these possibilities has been eliminated. A dosimetry experiment made at one of the major surveillance sites shows that the spectrum at that 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 wppm or less, which is inadequate for embrittlement. Copper and nickel impurities are shown to promote radiation strengthening at high doses but not at the low doses pertinent to the surveillance data. It is shown that a copper embrittlement scenario has other drawbacks, and it is argued that copper impurity is not responsible for the accelerated embrittlement of the HFIR surveillance specimens. The dosimetry experiment revealed unexpectedly high levels of reaction products in some of the fast flux monitors, which are found to be caused by an exceptionally high ratio of gamma ray flux to fast neutron flux at the pressure vessel. Gamma rays can also induce atomic displacements, leading to the suggestion that the accelerated embrittlement may be provoked by gamma irradiation.

A theoretical model has been constructed to calculate the time or fluence dependence of G-values for H2 production, G(H2), from the ion irradiation of the polymers polyethylene (PE), polypropylene (PP), polystyrene (PS), polycarbonate (PC), and Kapton. Measurements of the G(H2) for 1 Mev Ar+ over a fluence range from about 1 × 1011 to about 5 × 1013/mm2 have been made in order to determine the parameters of the model. The model is based upon rate equations describing the electronic-generation of and the interaction of a uniform distribution of free radicals. Satisfactory fits to the data could be made by adjusting two key parameters — the effective C-H bond energy and the hydrogen-carbon recombination rate constant relative to the hydrogen-hydrogen recombination rate constant. It was found that the effective C-H bond energy varied from the lowest value of ~ 8 eV for PE to the highest value of ~ 100 eV for Kapton. From the effective bond energy, an average value for hydrogen radical production, G(H·), was deduced. The effects of the parameters on the G-value versus time/fluence curves are shown and the significance of the parameters are discussed. The data was also compared to percolation model predictions, but the deviations between data and this model were seen to be large at high fluence.

D.J. Bacon ‘, D. Gavillet b, A. Horsewell ‘, S. Ishino d, L.K. Mansur +-, B.N. Singh C, H. Trinkaus f, H. Ullmaier f and M. Victoria b a Materials Sciences & Eng. Department, University of Liverpool, Liverpool L69 3BX, UK ’ Paul Scherrer Institut, CH-5232 Wigen PSL Switzerland ’ Materials Department, Rise National Laboratory, DK-4000 Roskilde, Denmark d Nuclear Engineering Department, University of Tokyo, Bunkyo-ku, Tokyo 113, Japan ’ Metals & Ceramics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA f Forschungszentrum Jiilich, IFF, D-52425 Jz’ilich, Germany

During irradiation of structural materials many processes take place simultaneously. These range from the elementary production of point defects, point defect clusters, and transmutation products to more complex processes of large scale cavity and precipitate formation, and alterations in macroscopic properties. Behavior that can be defined as transient or steady-state with respect to irradiation dose can often be recognized within each process. Transitions from one type of dose dependence to another can be particularly important. These transitions may signal changes in governing physical processes or they may result in changes in the performance of the material in technological applications. Transitions in dose dependence of point defect concentrations, helium buildup and related transient phenomena in swelling, irradiation creep, and embrittlement are discussed, with particular reference to austenitic Fe-Ni-Cr alloys. Special attention is given to the concept, derivation, and usefulness of irradiation variable shifts based on a requirement of invariance in point defect absorption. It is found that the doses at which transitions in dose dependence take place for swelling and creep for example, can vary by many orders of magnitude, depending upon irradiation conditions and material. Transitions for different phenomena can also be widely separated in dose. Under certain conditions, transient behavior can be the dominant feature in microstructure and property changes.

Eight complex alloys based on the composition Fe13Cr15Ni2Mo2Mn0.2Ti0.8Si0.06C were implanted simultaneously with 400 keV boron and 550 keV nitrogen, and investigated for microhardness changes and bending fatigue life. The dual implantation was found to decrease the fatigue life of all eight alloys although the implantation increased near-surface hardness of all eight alloys. This result was in contrast to the significant improvements found in the fatigue life of four B, N implanted simple Fe13Cr15Ni alloys. It was determined that the implantation suppressed surface slip band formation, the usual crack initiation site, but in the complex alloys, this suppression promoted a shift to grain boundary cracking. A similar phenomenon was also observed when the simple Fe13Cr15Ni alloys were simultaneously implanted with boron, nitrogen and carbon wherein fatigue life decreased, and gain, grain boundary cracks were observed. To test the hypothesis that ion implantation made the overall surface more fatigue resistant but led to a shift to grain boundary cracking, single crystal specimens of the ternary Fe15Cr15Ni were also implanted with boron and nitrogen ions. The fatigue life decreased for the single crystal specimens also, due to concentration of applied stress along fewer slip bands as compared to the control single crystal specimens were applied stress was relieved by slip band formation over the entire gauge region.

A workshop on Radiation Effects on Materials in High-Radiation Environments was held in Salt Lake City, Utah (USA) from August 13 to 15, 1990 under the auspices of the Division of Materials Sciences, Office of Basic Energy Sciences, US Department of Energy. The workshop focused on ceramics, alloys, and intermetallics and covered research needs and capabilities, recent experimental data, theory, and computer simulations. It was concluded that there is clearly a continuing scientific and technological need for fundamental knowledge on the underlying causes of radiation-induced property changes in materials. Furthermore, the success of many current and emerging nuclear-related technologies critically depend on renewed support for basic radiation-effects research, irradiation facilities, and training of scientists. The highlights of the workshop are reviewed and specific recommendations are made regarding research needs.

Essential aspects of the theory of radiation-induced swelling are reviewed. In particular, concepts central to the understanding of experimental data and the control of swelling by alloy design are discussed. The knowledge that a critical number of gas atoms is required in a cavity before point defect-driven swelling can begin is a most important contribution of theory. The mathematical expressions that have been derived for calculating this critical quantity are given in terms of materials parameters and irradiation conditions. After swelling begins, its magnitude as a function of dose is governed strongly by the relative sink strengths for point defects of dislocations and cavities, expressed in terms of an index of the microstructure, Q. High swelling and low swelling microstructures can be categorized into four types based on this index. Wide ranges of experimental swelling results covering ferritic/martensitic and austenitic alloys, neutron and ion irradiations and a variety of compositions and irradiation conditions are analyzed and found to be explained consistently within this framework. Based on the understanding gained, approaches to alloy design for swelling resistance are recommended.

Irradiation creep has been investigated in the Oak Ridge Research Reactor in an assembly spectrally tailored to achieve a He : dpa ratio of 12–14:1 appm/dpa in austenitic stainless steels. Temperatures of 60–400°C were investigated to address the requirements of near term fusion devices. It was found that austenitic alloys, especially PCA, have higher creep rates at 60°C than at 330 and 400°C. Since this phenomenon could not be explained by existing theoretical models, a new mechanism was proposed and a corresponding theoretical model was developed. Since vacancy migration times can be a few orders of magnitude longer than the irradiation times in this temperature regime, the immobile vacancies do not cancel climb produced by mobile interstitials absorbed at dislocations. The result is a high climb rate independent of stress-induced preferred absorption (SIPA) mechanisms. Preliminary calculations indicate that this mechanism coupled with preferred-absorption-driven glide at higher temperatures predicts a high creep rate at low temperatures and a weak temperature dependence of irradiation creep over the entire temperature range investigated.

Both low displacement rates and softened neutron spectrum favor survival of a higher fraction of point defects per displacement for producing micro-structural changes leading to hardening and embrittlement. Low displacement rate results in low bulk recombination rate. A high thermal to fast neutron flux ratio results in a large fraction of point defects produced in small cascades from (n,γ) and (n,α) reactions. Defects from such cascades generally avoid in-cascade recombination, while most of the defects created in large cascades produced by fast neutrons are lost to in-cascade recombination. Thus thermal neutrons produce more available defects per unit displacement dose. It is argued that the spectral effect may dominate the accelerated embrittlement observed in ferritic steels at the High Flux Isotope Reactor (HFIR) pressure vessel location. The rate effect is expected to be a secondary factor at temperatures as low as 50°C, where the HFIR data were obtained. Our analysis suggests generally that components subject to neutron environments with high thermal-to-fast ratios and irradiated at low temperatures may be subject to accelerated radiation effects.