Experimental verification of thermal decomposition of lead polonide

The chemical form of polonium in lead–bismuth eutectic (LBE) is an important issue, considering the problem of polonium contamination in nuclear systems that use LBE as a coolant and/or an irradiation target. It has been thought that polonium exists as lead polonide in LBE. Polonium forms compounds with several metals, some of which decompose at high temperatures. Thermal decomposition of lead polonide was not confirmed experimentally, but the temperature of decomposition was foreseen to be around 600 °C. In this paper, the thermal decomposition of lead polonide and its decomposition temperature were confirmed using neutron-irradiated LBE. Neutron-irradiated LBE ingots containing polonium-210 were heated at temperatures of 550 ± 10 °C or 630 ± 10 °C in a vacuum. Polonium, lead and bismuth evaporated from the LBE ingots, and were deposited onto the surface of type 316 stainless steel (316SS) plates at various controlled temperatures between 220 ± 20 °C and 450 ± 20 °C. After heating, the number of alpha particles emitted from polonium-210 deposited on 316SS plates was measured. The experimental results showed a clear difference in the alpha particle count rate, which indicated that lead polonide decomposed at a temperature between 550 ± 10 °C and 630 ± 10 °C.     

Polonium evaporation and adhesion experiments for the development of polonium filter in lead–bismuth cooled reactors

Fundamental experiments were performed to determine the adhesion characteristics of polonium to different metals and to develop a filter for polonium evaporated from neutron-irradiated LBE. The results of the first experiments suggested that adhesion characteristics are almost the same for stainless steel and nickel metal. The results of the preliminary experiments for a polonium filter suggested that stainless steel mesh with thin wires could effectively collect polonium evaporated from neutron-irradiated LBE. In the experiments, stainless steel wire mesh was used, but from the results of adhesion experiment, it is expected that the same effect can be obtained with wire mesh made of other kinds of metal.     

Polonium decontamination performance of stainless steel mesh filter for lead alloy-cooled reactors

Lead-Bismuth eutectic (LBE) has many good characteristics as a coolant for fast reactors. One of the issues remaining to be solved, however, is the polonium issue. The purpose of the present study was to estimate the decontamination performance of a polonium filter by experiment in the penetration condition. Two types of stainless steel wire meshes, fine wire mesh and loose wire mesh, were tested in the experiments. The results show that polonium filters made of stainless steel wire mesh can be very useful device for the removal of polonium in the gas phase. These filters can be used for the decontamination of primary loops by the baking method.     

Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 °C and 500 °C.During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 °C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 °C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.     

Development of polonium surface contamination measure in lead-bismuth eutectic coolant

Polonium contamination is one of the issues to be solved for use of lead-bismuth eutectic in a nuclear system. It is important to develop an effective method to remove polonium surface contamination. Some experiments were performed to remove polonium surface contamination by baking method. The results showed that characteristics are different between quartz glass contamination removal and stainless steel contamination removal. A method to estimate polonium density distribution in metal useful for investigation of polonium surface contamination was developed.     

Irradiation creep in deuteron bombarded stainless steel

A commercial and a high purity version of cold worked type 316 stainless steel was irradiated with 9 MeV deuterons at 300°C under tensile stresses between 100 and 350 MPa and the irradiation creep rate was measured. The results are qualitatively discussed in the light of present theoretical models.     

ADS散裂中子源材料不锈钢和钨的辐照效应(英文)

研究了80 MeV碳或85 MeV氟离子辐照在国产改进型316L不锈钢、普通不锈钢和钨中产生的辐照效应。实验结果表明,不锈钢的抗辐照性能比钨的好;它们中,国产改进型316L不锈钢具有最好的抗辐照性能。选用不锈钢做ADS散裂中子源的束窗等材料是较好的选择,采用国产改进型316L不锈钢是最佳的选择。     

Structural,electrical and magnetic measurements on oxide layers grown on 316L exposed to liquid lead–bismuth eutectic

Fast reactors and spallation neutron sources may use lead–bismuth eutectic (LBE) as a coolant. Its physical, chemical, and irradiation properties make it a safe coolant compared to Na cooled designs. However, LBE is a corrosive medium for most steels and container materials. The present study was performed to evaluate the corrosion behavior of the austenitic steel 316L (in two different delivery states). Detailed atomic force microscopy, magnetic force microscopy, conductive atomic force microscopy, and scanning transmission electron microscopy analyses have been performed on the oxide layers to get a better understanding of the corrosion and oxidation mechanisms of austenitic and ferritic/martensitic stainless steel exposed to LBE. The oxide scale formed on the annealed 316L material consisted of multiple layers with different compositions, structures, and properties. The innermost oxide layer maintained the grain structure of what used to be the bulk steel material and shows two phases, while the outermost oxide layer possessed a columnar grain structure.     

Corrosion studies in support of lead-bismuth cooled FBRs

The performance of structural materials in lead or lead-bismuth eutectic (LBE) systems is evaluated. The materials evaluated included several US steels (austenitic steel [316L], carbon steels [F-22, Fe-Si], ferritic/martensitic steels [HT-9 and 410]), and several experimental Fe-Si-Cr alloys that were expected to demonstrate corrosion resistance. The materials were exposed in either a dynamic corrosion cell for periods from 100 to 1,000 h at temperatures of 400, 500, 600 and 700°C, depending on material and exposure location. Weight change and optical SEM or X-ray analysis of the specimen were used to characterize oxide film thickness, corrosion depth, microstructure, and composition changes. The tests conducted with stainless steels (410, 316L and HT-9) produced mass transfer of elements (e.g., Ni and Cr) into the LBE, resulting in degradation of the material. With Fe-Si alloys a Si rich layer (as SiO₂) is formed on the surface during exposure to LBE from the selective dissolution of Fe.     

Emulation of neutron irradiation effects with protons: validation of principle

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 °C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 °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 °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 denuded zones were only observed in neutron-irradiated samples. No cavities were observed for either irradiating particle. For both irradiating particles, hardening increased with dose for both heats, showing a more rapid increase and approach to saturation for heat B. In normal oxygenated water chemistry (NWC) at 288 °C, stress corrosion cracking in the 304 alloy was first observed at about 1.0 dpa and increased with dose. The 316 alloy was remarkably resistant to IASCC for both particle types. In hydrogen treated, de-oxygenated water (HWC), proton-irradiated samples of the 304 alloy exhibited IG cracking at 1.0 dpa compared to about 3.0 dpa for neutron-irradiated samples, although differences in specimen geometry, test condition and test duration can account for this difference. Cracking in heat P in HWC occurred at about 5.0 dpa for both irradiating particles. Thus, in all aspects of radiation effects, including grain boundary microchemistry, dislocation loop microstructure, radiation hardening and SCC behavior, proton-irradiation results were in good agreement with neutron-irradiation results, providing validation of the premise that the totality of neutron-irradiation effects can be emulated by proton irradiation of appropriate energy.     

Hydrogen transport and solubility in 316L and 1.4914 steels for fusion reactor applications

Equations are given which describe the permeation rate, diffusivity and solubility of hydrogen over the range 250–600°C at pressures up to 10⁵Pa for the 316L stainless and modified 1.4914 martensitic candidate steels proposed for the construction of the Next European Torus (NET). For heat-treated 316L steel, the permeation rates measured agreed well with previous work and did not vary significantly from specimen to specimen or from batch to batch.

Measurements of the permeation rate of hydrogen and deuterium through the modified 1.4914 steel, believed to be the first made, show that the martensitic steel is significantly more permeable than the austenitic steel, by an order of magnitude at 250°C and a factor of five at 600°C. This difference could make it necessary to use permeation barriers on critical components made from the martensitic steel in order to reduce the tritium permeation rate to acceptable levels.     

Materials researches at the Paul Scherrer Institute for developing high power spallation targets

For the R&D of high power spallation targets, one of the key issues is understanding the behavior of structural materials in the severe irradiation environments in spallation targets. At PSI, several experiments have been conducted using the targets of the Swiss spallation neutron source (SINQ) for studying radiation damage effects induced by high energy protons and spallation neutrons. As well, experiments have been performed to investigate liquid lead-bismuth eutectic (LBE) corrosion and embrittlement effects on T91 steel under irradiation with 72 MeV protons. In this paper, an overview will be given showing a selection of results from these experiments, which include the mechanical properties and microstructure of ferritic/maretensitic (FM) steels (T91, F82H, Optifer etc.) and austenitic steels (EC316LN, SS 316L, JPCA etc.) irradiated to doses higher than ever attained by irradiation in a spallation environment, and the behaviors of T91 irradiated with 72 MeV protons in contact with flowing LBE.     

Effects of hold time and neutron irradiation on the low-cycle fatigue behaviour of type 316-CL and their consideration in a damage model

The creep fatigue behaviour of AISI type 316 L(N) plate material has been investigated in the temperature range of 450–750 °C by performing axial strain controlled tests with GRIM specimens. The creep and creep fatigue behaviour of austenitic stainless steel material is known to be prone to neutron irradiation-induced embrittlement. Therefore, the irradiation behaviour was studied by performing irradiation experiments in the High Flux Reactor (HFR) of Petten at 550 °C. A newly developed damage model for time-dependent damage was applied to describe the failure behaviour of AISI 316 L(N) in the cyclic tests performed.     

Investigation of polonium removal systems for lead-bismuth cooled fbrs

The alkaline extraction method is being investigated as a method to remove polonium (Po) from a lead-bismuth (Pb-Bi) cooled reactor. Tellurium (Te) was used as a surrogate for Po for these experiments to quantify the migration from Pb-Bi to molten alkaline (NaOH). Experiments provided direct evidence that molten NaOH can effectively remove Te from Pb-Bi. A reduction of three to four orders of magnitude in the Te concentration in the metals was measured in the experiments. The experimental results also showed a higher than expected concentration of NaOH in the Pb or Pb-Bi. A comparison between experiments operated at 427°C and 500°C indicates that the higher operating temperature produces a higher removal rate of Te. Experiments where hot NaOH was injected into the crucible at the same temperature as the lead-bismuth eutectic (LBE) resulted in the rapid removal rate of Te and remained constant until the experiment was completed.     

Deuterium Retention and Physical Sputtering of Low Activation Ferritic Steel

Low activation materials have to be developed toward fusion demonstration reactors. Ferritic steel, vanadium alloy and SiC/SiC composite are candidate materials of the first wall, vacuum vessel and blanket components, respectively. Although changes of mechanical-thermal properties owing to neutron irradiation have been investigated so far, there is little data for the plasma material interactions, such as fuel hydrogen retention and erosion. In the present study, deuterium retention and physical sputtering of low activation ferritic steel, F82H, were investigated by using deuterium ion irradiation apparatus. After a ferritic steel sample was irradiated by 1.7 keV D^ ions, the weight loss was measured to obtain the physical sputtering yield. The sputtering yield was 0.04, comparable to that of stainless steel. In order to obtain the retained amount of deuterium, technique of thermal desorption spectroscopy (TDS) was employed to the irradiated sample. The retained deuterium desorbed at temperature ranging from 450 K to 700 K, in the forms of DHO, D2, D2O and hydrocarbons. Hence, the deuterium retained can be reduced by baking with a relatively low temperature. The fiuence dependence of retained amount of deuterium was measured by changing the ion fiuence. In the ferritic steel without mechanical polish, the retained amount was large even when the fluence was low. In such a case, a large amount of deuterium was trapped in the surface oxide layer containing O and C. When the fluence was large, the thickness of surface oxide layer was reduced by the ion sputtering, and then the retained amount in the oxide layer decreased. In the case of a high fluence, the retained amount of deuterium became comparable to that of ferritic steel with mechanical polish or SS 316 L, and one order of magnitude smaller than that of graphite. When the ferritic steel is used, it is required to remove the surface oxide layer for reduction of fuel hydrogen retention. Ferritic steel sample was exposed to the environment of JFT-2M tokamak in JAERI and after that the deuterium retention was examined. The result was roughly the same as the case of deuterium ion irradiation experiment.     

Spectroscopic and microscopic investigation of the corrosion of D-9 stainless steel by lead-bismuth eutectic (LBE) at elevated temperatures. Initiation of thick oxide formation

Corrosion of 316/316L stainless steel by lead-bismuth eutectic (LBE) at elevated temperature was investigated by examination of samples after 1000, 2000, and 3000 h of exposure at 550 °C, using SEM, XPS with sputter depth profiling, and TEM. The process by which localized oxide failure becomes extensive thick oxide formation was investigated. Under our experimental conditions, iron was observed to migrate outward while chromium did not migrate above the original metal surface. The thin oxide layer on the D-9 sample resembled 316L cold-rolled samples, while the thick oxide on D-9 resembled annealed 316L oxide. With continued exposure, thick oxide grew to cover the entire surface.     

Compatibility of martensitic/austenitic steel welds with liquid lead bismuth eutectic environment

The high-chromium ferritic/martensitic steel T91 and the austenitic stainless steel 316L are to be used in contact with liquid lead-bismuth eutectic (LBE), under high irradiation doses. Both tungsten inert gas (TIG) and electron beam (EB) T91/316L welds have been examined by means of metallography, scanning electron microscopy (SEM-EDX), Vickers hardness measurements and tensile testing both in inert gas and in LBE. Although the T91/316L TIG weld has very good mechanical properties when tested in air, its properties decline sharply when tested in LBE. This degradation in mechanical properties is attributed to the liquid metal embrittlement of the 309 buttering used in TIG welding of T91/316L welds. In contrast to mixed T91/316L TIG welding, the mixed T91/316L EB weld was performed without buttering. The mechanical behaviour of the T91/316L EB weld was very good in air after post weld heat treatment but deteriorated when tested in LBE.     

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

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

Alkaline extraction of polonium from liquid lead bismuth eutectic

The production of highly radiotoxic polonium isotopes poses serious safety concerns for the development of future nuclear systems cooled by lead bismuth eutectic (LBE). In this paper it is shown that polonium can be extracted efficiently from LBE using a mixture of alkaline metal hydroxides (NaOH + KOH) in a temperature range between 180 and 350 °C. The extraction ratio was analyzed for different temperatures, gas blankets and phase ratios. A strong dependence of the extraction performance on the redox properties of the cover gas was found. While hydrogen facilitates the removal of polonium, oxygen has a negative influence on the extraction. These findings open new possibilities to back up the safety of future LBE based nuclear facilities.