Uncertainty Analyses of Neutron Cross Sections for 235U in the Resonance Region

The helium embrittlement behavior of 316FR austenitic stainless steel was investigated by a tensile test at 750°C using miniature tensile specimens, which were helium-implanted below 100°C up to 5, 30, and 100 appm using a cyclotron accelerator, and were post-implantation-annealed at 750°C for 10 and 100 h. The helium-implanted specimens showed a fully intergranular fracture regardless of the helium concentration and annealing time. No microstructural changes in the as-implanted specimen up to 30 appm and formation of a small number of helium bubbles due to the post-implantation annealing were observed. The gradual release of the helium during the tensile test started after the yielding, and a sharp peak of the helium release was detected in the final fracture phase. The total number of helium atoms released was strongly dependent on the implanted helium concentration, rather than on the annealing time.     

Mechanical properties and microstructure of helium-implanted beryllium

0.2 mm thick specimens of beryllium have been homogeneously implanted with helium. Implantation temperatures ranged from 100 to 600 °C, and final helium concentrations from 30 to 800 appm. Tensile tests at temperatures between 20 and 600 °C were carried out with testing temperature both equal to and lower than the implantation temperatures. For practicality all conditions of helium-implanted specimens, ductility decreased and yield and ultimate tensile strength increased as compared to the unimplanted specimens. The amount of embrittlement and strengthening, however, depended sensitively on implantation dose, implantation temperature, and tensile test temperature. The formation of helium bubbles, dislocation loops, and dislocation networks and the fracture modes were observed by transmission and scanning electron microscopy, respectively. Two ranges of embrittlement can be distinguished. They are attributed to different mechanisms: matrix strengthening is the dominant mechanism at low temperatures, and loss of grain boundary cohesion at high temperatures. It is concluded that in both temperature regimes the embrittlement is dominated by helium and not by the displacement defects introduced by its implantation.     

Recent applications of small-angle neutron scattering in the characterization of irradiated steels for nuclear technologies

Small-angle neutron scattering (SANS) is a powerful experimental tool to investigate the microstructural evolution under irradiation in steels for fission and future fusion reactor systems. We present recent SANS results concerning the modelling of helium bubble growth in F82H-mod. steel implanted with α-particles and the dose dependence of microstructural radiation damage in Eurofer-97 steel for fusion reactors irradiated at 250 °C. The discussion of these results is focussed on the quality of the metallurgical information obtained by such SANS measurements and consequently on their usefulness also for engineering and design purposes.     

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

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

Post irradiation plastic properties of F82H derived from the instrumented tensile tests

F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 °C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 °C.     

Tensile fracture characteristics of nanostructured ferritic alloy 14YWT

High temperature tensile fracture behavior has been characterized for the nanostructured ferritic alloy 14YWT (SM10 heat). Uniaxial tensile tests were performed at temperatures ranging from room temperature to 1000 °C in vacuum at a nominal strain rate of 10⁻³ s⁻¹. Comparing with the existing oxide dispersion strengthened (ODS) steels such as Eurofer 97 and PM2000, the nanostructured alloy showed much higher yield and tensile strength, but with lower elongation. Microstructural characterization for the tested specimens was focused on the details of fracture morphology and mechanism to provide a feedback for process improvement. Below 600 °C, the fracture surfaces exhibited a quasi-brittle behavior presented by a mixture of dimples and cleavage facets. At or above 600 °C, however, the fracture surfaces were fully covered with fine dimples. Above 700 °C dimple formation occurred by sliding and decohesion of grain boundaries. It was notable that numerous microcracks were observed on the side surface of broken specimens. Formation of these microcracks is believed to be the main origin of the poor ductility of 14YWT alloy. It is suggested that a grain boundary strengthening measure is essential to improve the fracture property of the alloy.     

Fracture behavior of austenitic stainless steels irradiated in PWR

Fracture behavior of cold-worked 316 stainless steels irradiated up to 73 dpa in a pressurized water reactor was investigated by impact testing at −196, 30 and 150 °C, and by conventional tensile and slow tensile testing at 30 and 320 °C. In impact tests, brittle IG mode was dominant at −196 °C at doses higher than 11 dpa accompanying significant decrease in absorbed energy. The mixed IG mode, which was characterized by isolated grain facets in ductile dimples, appeared at 30 and 150 °C whereas the fracture occurred macroscopically in a ductile manner. The sensitivity to IG or mixed IG mode was more pronounced for higher dose and lower test temperature. In uniaxial tensile tests, IG mode at a slow strain rate appeared only at 320 °C whereas mixed IG mode appeared at both 30 and 320 °C at a fast strain rate. A compilation of the results and literature data suggested that IG fracture exists in two different conditions, low-temperature high-strain-rate (LTHR) and high-temperature low-strain-rate (HTLR) conditions. These two conditions for IG fracture likely correspond to two different deformation modes, twining and channeling.     

Hoop strength and ductility evaluation of irradiated fuel cladding

The objective of this study is to evaluate the hoop-directional mechanical properties comprising strength such as yield strength and ultimate tensile strength as well as mechanical ductility such as uniform elongation and total elongation. Therefore, in this paper, the ring tensile tests were performed in order to evaluate the mechanical properties of high burn-up fuel cladding under a hoop loading condition in a hot cell. The tests were performed with Zircaloy-4 nuclear fuel cladding whose burn-up is approximately 65,000 MWd/tU in the temperature range of room temperature to 800 °C. All the experiments were carried out at a constant strain rate of 0.01/s.On the basis of the ring tensile tests for a high burn-up Zircalay-4 cladding, the following conclusions were drawn. Firstly, the mechanical properties are abruptly degraded beyond 600 °C, which corresponds to a design-basis accident condition such as a RIA. Secondly, the un-irradiated fuel cladding showed ductile fracture behaviors such as 45° shear type fracture, cup and cone type fracture, cup and cup type fracture and chisel edge type fracture. While the high burn-up Zircalay-4 cladding showed a brittle fracture behavior even at the high temperatures (e.g. over 600 °C) which are achievable during a RIA. Thirdly, in the case of the high burn-up Zircalay-4 cladding, the strength, ductility and the energy to break are strongly dependent on the material property itself which are degraded by oxidation and hydriding during an operation rather than the temperature. Fourthly, hydride rim formation in the vicinity of metal-oxide interface can play an important role in the degradation of the mechanical properties for high burn-up fuel cladding.     

The effects of large concentrations of helium on the mechanical properties of neutron-irradiated stainless steel

Tensile and creep properties have been determined on specimens of type 316 stainless steel irradiated in the High Flux Isotope Reactor in the range 380 to 785°C. Irradiation of type 316 in this reactor partially simulates fusion reactor irradiation, with displacement damage levels up to 120 dpa and helium contents up to 6000 appm achieved in two years. Samples irradiated in the annealed condition to about 100 dpa and 4000 appm helium showed an increased yield strength between 350 and 600°C and, except at 350°C, a reduced ultimate tensile strength compared with values for the unirradiated material. Samples irradiated in the 20%-cold-worked condition showed decreases in both yield and ultimate tensile strengths at all test temperatures. The irradiated samples of both annealed and cold-worked material exhibited little strain hardening, and total elongations were small and became zero,for tests at 650° C. Tensile tests at 575°C and creep-rupture tests at 550°C showed strong effects of fluence on strength and ductility for helium contents above about 30 appm. Optical metallography showed extensive carbide precipitation at all temperatures and precipitation of a second phase, believed to be sigma, at the higher temperatures.     

Evaluation of effect of hydrogen on toughness of Zircaloy-2 by instrumented drop weight impact testing

Hydride-assisted degradation in fracture toughness of Zircaloy-2 was evaluated by carrying out instrumented drop-weight tests on curved Charpy specimens fabricated from virgin pressure tube. Samples were charged to 60 ppm and 225 ppm hydrogen. Ductile-to-brittle-transition behaviour was exhibited by as-received and hydrided samples. The onset of ductile-to-brittle-transition was at about 130 °C for hydrided samples, irrespective of their hydrogen content. Dynamic fracture toughness (K_ID) was estimated based on linear elastic fracture mechanics (LEFM) approach. For fractures occurring after general yielding, the fracture toughness was derived based on equivalent energy criterion. Results are supplemented with fractography. This simple procedure of impact testing appears to be promising for monitoring service-induced degradation in fracture toughness of pressure tubes.     

Evaluation of grain boundary embrittlement of phosphorus added F82H steel by SSTT

Non-hardening embrittlement (NHE) can be happened by a large amount of He on grain boundaries over 500–700 appm of bulk He without hardening at fusion reactor condition. Especially, at high irradiation temperatures (>≈420 °C), NHE accompanied by intergranular fracture affects the severe accident and the safety of fusion blanket system. Small specimen tests to evaluate fracture toughness and Charpy impact properties were carried out for F82H steels with different levels of phosphorous addition in order to simulate the effects of NHE on the shift of transition curve. It was found that the ductile to brittle transition temperature (DBTT) and reference temperature (T₀) after phosphorous addition is shifted to higher temperatures and accompanied by intergranular fracture at transition temperatures region. The master curve approach for evaluation of fracture toughness change by the degradation of grain boundary strength was carried out by referring to the ASTM E1921.     

Gaseous hydrogen embrittlement of PH 13-8 Mo steel

In this study, notched tensile and fatigue crack growth tests in gaseous hydrogen were performed on PH 13-8 Mo stainless steel specimens at room temperature. These specimens were susceptible to hydrogen embrittlement (HE), but at different degrees, depending on the aging conditions or the microstructures of the alloys. In hydrogen, the accelerated fatigue crack growth rate (FCGR) usually accompanied a reduced notched tensile strength (NTS) of the specimens, i.e., the faster the FCGR the lower the NTS. It was proposed that the same fracture mechanism could be applied to these two different types of specimens, regardless of the loading conditions. Rapid fatigue crack growth and high NTS loss were found in the H800 (426 °C under-aged) and H900 (482 °C peak-aged) specimens. The HE susceptibility of the steel was reduced by increasing the aging temperature above 593 °C, which was attributed to the increased amount of austenite in the structure. Extensive quasi-cleavage fracture was observed for the specimens that were deteriorated severely by HE.     

Effect of Pre-Hydriding on Post-Irradiation Ring-Tensile Properties of Zircaloy-2 Cladding Tubes

Abstract

Zircaloy-2 tubes were hydrided up to a nominal content of 200 ppm and irradiated as fuel claddings in HBWR. Post-irradiation ring-tensile testing revealed that hydrogen enhances the irradiation-induced decrease of elongation and wall thickness reduction at room temperature. On the other hand, no effect of hydrogen was observed on ultimate tensile strength. With testings at 300°C, the effect was negligible on elongation too. From the evaluation of the test results including metallographic observation of ring specimens after fracture, it was concluded that segregation of hydrides due to thermal diffusion of hydrogen during irradiation was at least a part responsible to the above effect of hydrogen enhancing embrittlement.     

Tensile behaviour of 9Cr-1Mo tempered martensitic steels irradiated up to 20 dpa in a spallation environment

Tensile specimens of 9Cr-1Mo (EM10) and mod 9Cr-1Mo (T91) martensitic steels in the normalized and tempered metallurgical conditions were irradiated with high energy protons and neutrons up to 20 dpa at average temperatures up to about 360 °C. Tensile tests were carried out at room temperature and 250 °C and a few samples were tested at 350 °C. The fracture surfaces of selected specimens were characterized by Scanning Electron Microscopy (SEM). While all irradiated specimens displayed at room temperature considerable hardening and loss of ductility, those irradiated to doses above approximately 16 dpa exhibited a fully brittle behaviour and the SEM observations revealed significant amounts of intergranular fracture. Helium accumulation, up to about 0.18 at.% in the specimens irradiated to 20 dpa, is believed to be one of the main factors which triggered the brittle behaviour and intergranular fracture mode. One EM10 and one T91 specimen irradiated to 20 dpa were annealed at 700 °C for 1 h following irradiation and subsequently tensile tested. In both cases, a remarkable recovery of ductility and strain-hardening capacity was observed after annealing, while the strength remained significantly above that of the unirradiated material.     

Synergistic effects of implanted helium and hydrogen and the effect of irradiation temperature on the microstructure of SiC/SiC composites

The microstructure of near-stoichiometric fiber SiC/SiC composites implanted with He and H ions was studied at implantation temperatures of 1000 and 1300 °C. The average size of He bubbles in the CVI SiC matrix decreases with increasing concentration of implanted H ions. Moreover, the number density of He bubbles increases with increasing irradiation temperature and amount of implanted H. At the irradiation temperature of 1000 °C, He bubbles were mainly formed at grain boundary within the matrix. On the other hand, He bubbles were formed both at grain boundaries and within grains at the irradiation temperature of 1300 °C. The average size of He bubbles at grain boundaries was much larger than within the grain. The average size of He bubbles in the fiber was smaller than that in the matrix in all cases.     

Annealing of a cold-rolled Pu/1 wt% Ga alloy

The variation of the yield strength, strain hardening, and ductility during annealing of cold-rolled Pu/1 wt% Ga alloy has been determined by means of tensile tests at room temperature. Specimens rolled 20, 40, 65 and 80% were annealed at temperatures between 200 and 300 °C for times up to 1000 min. It was found that there was a linear relation between the strain hardening exponent and the fractional decrease in yield strength, and that the relation was independent of the prior deformation in the specimens or temperature of annealing. As a result, the variation of the ultimate strength with the yield strength differed for different amounts of prior cold work. Fracture elongations varied approximately linearly with the yield strength in a manner independent of prior deformation or annealing temperature. The final softening fraction by recovery determined from the yield strength of 20% rolled specimens varied from 0.1 at 200 °C to 0.5 at 300 °C. The final softening fraction by recovery did not vary with the amount of prior cold work in specimens annealed at 200 °C.     

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between −160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between −170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.     

Deformation and fracture properties in neutron irradiated pure Mo and Mo alloys

The effect of neutron irradiation on the mechanical properties of select molybdenum materials, unalloyed low carbon arc-cast (LCAC) Mo, Mo-0.5% Ti-0.1% Zr (TZM) alloy, and oxide dispersion-strengthened (ODS) Mo alloy, was characterized by analyzing the temperature dependence of mechanical properties. This study assembles the tensile test data obtained through multiple irradiation and post-irradiation experiments, in which tensile specimens were irradiated up to 13.1 dpa at 80-1000 °C and tested at −194 to 1000 °C. Irradiation at 80-609 °C increased yield stress significantly, up to 170%, while the increase of yield stress after irradiation at 784-936 °C was not significant. The plastic instability stress was strongly dependent on test temperature but was nearly independent of irradiation dose and temperature. The true fracture stress showed weak dependences on test temperature, irradiation dose and temperature when ductile failure occurred. Among the test materials the stress-relieved ODS material in the longitudinal direction (ODS-LSR) displayed the highest resistance to irradiation embrittlement due to its relatively high fracture stress. The critical temperature for shear failure (CTSF) was defined and evaluated for the test materials and the CTSF values were compared with the ductile-to-brittle transition temperatures (DBTT) based on ductility data.     

A study of grain boundary sliding in copper with and without an addition of phosphorus

Copper will be used as a corrosion barrier in the storage of high level nuclear waste. In order to improve the creep fracture properties of the material it will contain 30-50 ppm of phosphorus, OFP copper as opposed to OF copper without P. It has been suggested that the phosphorus impedes grain boundary sliding in copper and recently a quantitative theory based on this idea has shown that there is no risk for creep-brittle fracture of OFP copper under waste storage conditions. In order to verify the basis of this theory grain boundary sliding has been investigated in copper with and without a P addition. The method has been to examine intentionally scratched surfaces of tensile specimens tension tested to plastic strains of 1%, 2% and 4% at 150 and 200 °C. After testing specimen surfaces have been examined in SEM and sliding distances have been measured as in-surface displacement of scratches. The results have been plotted as distribution functions where the fraction of slides smaller than a given value is plotted versus sliding distance. The result is that in most cases the distribution functions for OF and OFP copper overlap. In a small number of cases there is a tendency that less sliding has occurred in OFP copper. The overall conclusion is however that although there may be a slight difference between the materials with regard to grain boundary sliding it is not large enough to explain the observed difference in creep brittleness. Tension tests to fracture in the temperature range 100-200 °C show that the tensile properties of the two copper qualities are more or less identical until intergranular cracking starts in the OF copper. Then the flow stress decreases in comparison with OFP. It is suggested that at least part of the observed differences in creep strength between the two coppers may be due to the effect of intergranular cracking.     

Liquid Pb-Bi embrittlement effects on the T91 steel after different heat treatments

The present work aims to investigate the susceptibility of ferritic/martensitic steels of different strength to the embrittlement of liquid Pb-Bi eutectic (LBE). Slow strain rate tensile (SSRT) tests on specimens of the T91 steel in three tempering conditions at 500, 600 and 760 °C were conducted in Ar and in LBE at temperatures between 150 and 500 °C. For the specimens tempered at 760 °C (the normal tempering condition) the susceptibility of the steel to LBE embrittlement appeared at temperatures between 300 and 450 °C. With increasing the strength of specimens by lowering the tempering temperature, specimens tempered at 600 and 500 °C demonstrated more pronounced embrittlement effects, reflected by wider and deeper ‘ductility-troughs’. The results suggest that ferritic/martensitic steels with higher strength are more susceptible to LBE embrittlement. The LBE embrittlement effects can be attributed to the decrease of fracture stress resulted from the ‘weakening inter-atomic bond’ by LBE contacting at crack tips.