Unirradiated State Research Articles

Microstructural and Mechanical Characterization of Radiation Effects in Model Reactor Pressure Vessel Steels

Abstract This paper presents results of radiation-induced nanoscale microstructural changes measured by Small-Angle Neutron Scattering (SANS) on ASTM-type reactor pressure vessel (RPV) steels. Five different base metals and one weld metal were investigated. The irradiation was performed in a VVER-type power reactor to three different dose levels, maximally 0.14 dpa, at an irradiation temperature of 255°C. The volume content of the micostructural features measured is correlated with the irradiation hardening and embrittlement. In every case clear radiation-related hardening, embrittlement, and microstrucural effects are proven. Radiation produces nanoscale scattering defects of a radius between 0 < R < 3 nm with the maximum near R = 1 nm. Fluence and chemical composition do not or hardly change the shape of the size distribution but strongly influence the volume fraction. The radiation sensitivity is mainly influenced by the copper content. Copper also increases the A-ratio. There are good correlations between volume fraction of the nanoscale radiation defects and the radiation hardening or embrittlement. Annealing at 475°C produces almost full recovery of the mechanical properties. However, the original microstructure of the unirradiated state is not completely reproduced.

Radiation-induced defects in MOS structures after irradiation with high-energy Ar, Kr, Bi heavy ions

Radiation damage in the MOS (metal-oxide-semiconductor) structures irradiated with Ar, Kr and Bi heavy ions with energies from 1 to 7.5 MeV/nucl. has been studied using charge deep level transient spectroscopy (Q-DLTS) and feedback charge capacitance voltage ( C – V ) techniques. Four dominant well-known deep levels such as vacancy–oxygen (VO) complex, divacancy (VV 2−) in the double negative charge state and composite VV −+VSb complex were registered versus ion fluence and annealing temperature. It was shown that VO center concentration at the near-surface silicon substrate layer is determined by ion energy loss in elastic collisions and no effect of high-level electronic stopping power was detected. At the same time, the value of the positive radiation-induced defect charge in the oxide layer was decreased under Bi ions irradiation. Thermal treatment in air ambient at 250 °C restores CV characteristics of the MOS samples to that of un-irradiated state.

Microstructure and Bend Ductility of W-0.3 mass%TiC Alloys Fabricated by Advanced Powder-Metallurgical Processing

Low temperature embrittlement, recrystallization embrittlement and radiation embrittlement in tungsten and its alloys are critical issues for use as high heat flux components and high-power density structural materials. In order to establish a process for microstructural control to improve the resistance to such embrittlement, modification of powder-metallurgical processing is proposed to avoid three microstructural factors giving detrimental effects on the ductility: (1) precipitation of the brittle W2C phase, (2) heterogeneity in grain size and particle distributions, and (3) loss of carbon which is a constituent of transition metal carbides. The processing was applied to fabricate W-0.3 mass%TiC alloys with a microstructure of fine grains and nano-sized dispersoids of TiC. Transmission electron microstructural observations and three-point bending tests at room temperature were performed on the alloys in the unirradiated state. It is demonstrated that the developed alloys are almost free from the three microstructural factors and exhibit appreciable room-temperature ductility before fracture in the as-forged and as-rolled states, but not in the as-HIPed state. This beneficial effect of plastic working on ductility improvement is strongly dependent on grain size and becomes prominent with decreasing grain size. Success in fabricating consolidated bodies with grain sizes as small as 0:17� 0:4 mm and a high relative density of around 99% is presented.

Interaction of deoxyribonucleic acid with H1 histone in the unirradiated and gamma irradiated states in vitro

This study was undertaken in order to find out the effect of solvent composition on mode of interaction of H 1 histone with DNA, suitability of DNA, H 1 histone and DNA-H 1 complex as in vitro biological dosimeters and protection possibility of DNA by H 1 histone against ionizing radiation-induced damage. Analytical techniques were used for the study of complex formation. Gamma irradiation was used to deliver doses to the biopolymers and their complexes. Equilibrium dialysis method was used for complex formation. FBX dosimetry was done for the delivery of total dose and dose rate determination. DNA : H 1 complex formation ratio were : 4.17 x 10 - 9 : 4.652 x 10 - 6 ML - 1 in 10 - 3 M phosphate buffer; 4.17 x 10 - 9 : 1.163 x 10 - 6 ML - 1 in 0.1 M phosphate buffer and 2.085 x 10 - 9 : 9.304 x 10 - 7 ML - 1 in 0.15 M SSC buffer. Linearity in response with absorbed dose up to 50 Gy (dose rate 1.50 Gy/min) was observed with (i) DNA at 207 nm and 259 mn at 4.17 x 10 - 9 ML - 1 , (ii) with H 1 histone at 202 nm and 270 nm at 1.163 x 10 - 6 ML - 1 in 0.1 M phosphate buffer, pH 7.0, (iii) with DNA: H 1 at 206 nm and 260 nm in 0.1 M phosphate buffer, (iv) with DNA: H 1 at 226 nm and 260 nm in SSC buffer. A constant decrease in fluorescent intensity (Ex 280 nm, Em 338 nm) was observed for H 1 histone with increase in dose. The average percentage hyperchromicity in 0.1 M phosphate buffer was 20% for DNA, 11% for H 1 histone and 17% for DNA: H 1 complex. In 0.15 M SSC buffer, for the complex it was 14% for a dose of 50 Gy at a dose rate of 1.5 Gy/min. The degrees of unfolding of the chromophores for this complex with dose were different in different solvents. Complex formations are due to electrostatic, minor groove binding and intercalation modes and are solvent dependent. DNA, H 1 and DNA: H 1 complex can be used as in vitro biological dosimetes. Binding modes are responsible for protection of DNA by H 1 histone up to a dose of 50 Gy.

Small-angle neutron scattering study on the effect of hydrogen in irradiated reactor pressure vessel steels

Hydrogen uptake can enhance the neutron embrittlement of reactor pressure vessel (RPV) steels. This suggests that irradiation defects act as hydrogen traps. The evidence of hydrogen trapping was investigated using the small-angle neutron scattering (SANS) method on four RPV steels. The samples were examined in the unirradiated and irradiated states and both in the as-received condition and after hydrogen charging. Despite the low bulk content of hydrogen achieved after charging with low current densities, an enrichment of hydrogen in small microstructural defects could be identified. Preferential traps were microstructural defects in the size range of ≈>10nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.

A unified model to describe the anisotropic viscoplastic behaviour of zircaloy-4 cladding tubes between 320 °C and 400 °C from an unirradiated state to high fluence and under fast neutron flux

This paper proposes a damaged viscoplastic model to simulate, for different isotherms (320, 350, 380, 400 and 420°C), the out-of-flux anisotropic mechanical behavior of Zircaloy-4 claddings in a stress relieved state over the fluence range 0-85.10 24 nm -2 (E>lMeV). The model, identified from uni and biaxial tests conducted at 350 and 400°C, is validated from tests made at 320, 380 and 420°C. This model is able to simulate strain hardening under an internal pressure followed by a stress relaxation period, since the loading produces an interaction between the pellet and the cladding (thermal creep). Both the integration of a scalar state variable, characterizing the damage caused by a bombardment with neutrons, and the modification of the static recovery law allowed us to simulate the fast neutron flux effect (irradiation creep).

Radiation damage effects in superplastic 3Y-TZP irradiated with Zr ions

The 3 mol% yttria stabilized tetragonal zirconia polycrystals (3Y-TZP) were irradiated using 130 MeV Zr 11+ ions in the TANDEM accelerator facility at Tokai Research Establishment, JAERI. Irradiation was performed with the fluence of 3.5 × 10 12 and 2.1 × 10 13 ions/cm 2. Residual stresses and changes in mechanical properties caused by the ion irradiation and the effects of the subsequent annealing are studied. The occurrence of compressive residual stress and increases in hardness and fracture toughness were found at the surface regions of as-irradiated specimens. It was found from the subsequent annealing that these quantities were decreased gradually with raising the annealing temperature and returned to those of un-irradiated state at around 1173 K. A most probable cause of the increases in the hardness and fracture toughness after the irradiation may, therefore, be the residual compressive stress left in the irradiated surface region.

Effect of neutron irradiation on low-cycle fatigue of GlidCopAl25IG alloy for ITER applications

With the ITER in operation, one of the factors determining the life time of high-heat flux components, such as the divertor and first wall, will be the materials resistance to cyclic loading. The report presents the results of an investigation of low cycle fatigue (LCF) of GlidCopAl25IG (CR+ann) alloy and GlidCopAl25IG (HIP) from GlidCopAl25IG/316LN HIP joints in the unirradiated state and after neutron irradiation up to 0.4 dpa at T irr=150 or 300 °C. Testing of base alloy specimens irradiated at 150 °C demonstrates that on the whole the number of cycles to fracture is decreased. When tested at 300 °C, the specimens irradiated at 300 °C up to 0.4 dpa demonstrate a slight increase in the number of cycles to fracture as compared with the unirradiated ones.

Lifetime of Irradiated Iter Divertor Heat Sink

Comparative lifetime analysis of the ITER divertor Plasma Facing Component (PFCs) - Carbon Fiber Composite (CFC) monoblock with a heat sink tube made from CuCrZr bronze has been performed. The aim of this analysis was to evaluate the irradiation effects in CFC and CuCrZr on the monoblock lifetime. The decreasing of CFC thermal conductivity and changing of CuCrZr mechanical properties such as tensile and yield strength and elongation during the monoblock life were taken into consideration. Cumulative fatigue damage for both irradiated and unirradiated states of the heat sink tube was evaluated.

Determination of Reference Temperature T 0 for Steel JRQ in an Unirradiated State and Construction of a Master Curve

This paper considers the results of testing reactor steel JRQ involving determination of the fracture toughness and the reference temperature T0, construction of a Master Curve, and evaluation of the boundaries of regions with different failure probability. The tests were conducted on small specimens (1/2T) made from blocks of steel JRQ supplied by the IAEA within the framework of the Round Robin program.

A simple paper chromatographic technique for detection of different types of histones in the unirradiated and gamma irradiated states in vitro

Paper chromatographic technique has been employed to detect low levels (20-25 μg/ml) of H 1 , H 3 and H 4 histones under physiological pH range in the native, heat denatured and gamma irradiated states (10-50 Gy). The R t values obtained for these histones under different experimental conditions have rendered insight about the configurational alterations.

Influence of neutron-irradiation-induced defects on the flux pinning inHgBa2Ca2Cu3O8+xsingle crystals

The influence of fast neutron irradiation on flux pinning in ${\mathrm{HgBa}}_{2}{\mathrm{Ca}}_{2}{\mathrm{Cu}}_{3}{\mathrm{O}}_{8+x}$ single crystals ${(T}_{c}=120\mathrm{K})$ subjected to a fluence of $5\ifmmode\times\else\texttimes\fi{}{10}^{17}{\mathrm{cm}}^{\mathrm{\ensuremath{-}}2}$ was studied. Magnetic measurements were performed using a commercial superconducting quantum interference device magnetometer and a miniaturized torque magnetometer. In the unirradiated state, the irreversibility line (IL), plotted as $\mathrm{ln}{(H}_{\mathrm{irr}})$ vs $\mathrm{ln}(1\ensuremath{-}{T}_{\mathrm{irr}}{/T}_{c}),$ shows two slopes. At higher temperatures (85--100 K) the IL is described by a power-law dependence ${H}_{\mathrm{irr}}{(T)=H}_{\mathrm{irr}}(0)$ $(1\ensuremath{-}{T}_{\mathrm{irr}}{/T}_{c}{)}^{\ensuremath{\alpha}}$ with $\ensuremath{\alpha}\ensuremath{\approx}2.1.$ At lower temperatures (25--60 K), a more rapid change of ${H}_{\mathrm{irr}}$ with temperature is observed, with the exponent $\ensuremath{\alpha}\ensuremath{\approx}4.8.$ Irradiation shifts the IL to significantly higher magnetic fields/temperatures, where it is rather well described by a single power-law dependence with the exponent $\ensuremath{\alpha}\ensuremath{\approx}2.3.$ The effective mass anisotropy $\ensuremath{\gamma}{=(m}_{c}{/m}_{\mathrm{ab}}{)}^{1/2},$ as determined from torque measurements, decreases after neutron irradiation. The shielding current density as a function of temperature up to 60 K is well approximated by the exponential dependence ${j}_{s}{(T)=j}_{s}(0)\mathrm{exp}(\ensuremath{-}{T/T}_{0}).$ Irradiation increases the characteristic temperature ${T}_{0}$ from about 5.9 K (in the as-prepared crystal) to ${T}_{0}=9.4\mathrm{K},$ clearly reflecting a slower decay of ${j}_{s}$ with temperature. Neutron-generated defects significantly increase ${j}_{s}$ and suppress the ``fishtail effect'' (an increase of ${j}_{s}$ with magnetic field), which was present for the unirradiated crystal.

The influence of neutron irradiation on the fatigue performance of OFHC copper and a dispersion strengthened copper alloy

The fatigue performance of pure copper of the oxygen free, high conductivity (OFHC) grade and a dispersion strengthened copper alloy, Glid Cop TM CuAl-25 was examined with and without irradiation exposure. Mechanical testing was carried out to establish the fatigue lives of these materials in the unirradiated and irradiated states. Fatigue specimens of these two materials were irradiated with fission neutrons in the DR-3 reactor at Risø with a flux of ≈2.5 × 10 −17 n/m 2 s ( E > 1 MeV) to fluence levels of 1.5–2.5 × 10 24 n/m 2 ( E > 1 MeV) at ≈47 and 100°C. Specimens irradiated at 47°C were fatigue tested at room temperature, whereas those irradiated at 100°C were tested at the irradiation temperature. These investigations demonstrated that, while irradiation causes significant hardening of the materials, the hardening appears to have no negative impact on fatigue performance. In fact, the fatigue performance of the CuAl-25 alloy is considerably better in the irradiated than that in the unirradiated state tested both at 22°C and 100°C. Microstructural observations of the fatigue microstructures and the fracture surfaces are useful in understanding this conclusion.

Irradiation-induced low-temperature creep of ds copper alloy

Abstract The data are presented on in-pile creep of dispersion strengthened (DS) copper alloys MAGT 0.2 (Cu-0.2Al2O3). The experiment was performed in the SM-2 reactor in the Core position at T irr ∼ 80°C and a dose of 5dpa. Pipe samples loaded by inner helium pressure were irradiated for 41.2 days. Measurements of the pipe diameter after irradiation demonstrated that a low-temperature creep is observed in the DS alloy MAGT 0.2, the rate of which at a hoop stress of 117MPa is as high as 2 × 10−9s−1. The results of TEM investigations of the irradiated specimens structure are presented. The analysis of the obtained results and their comparison with the available data on creep of copper alloys in the unirradiated state allowed for the conclusion that irradiation inereases essentially the creep of DS copper alloys. This effect is to be taken into account when estimating the life time of the ITER components and units.

Thermal annealing and optical darkening effects in CsI(Tl) crystals

The effects of thermal annealing and optical darkening on Φ2.54×2.54 cm 2 CsI(Tl) crystals damaged by exposure to 60 Co gamma radiation and on unirradiated crystals have been studied. Thermal annealing can restore radiation-damaged crystals to their unirradiated state and can even improve the scintillation performance of unirradiated crystals. Optical darkening of crystals with an intense UV source leads to the production of a number of bands in the absorption spectra of crystals, which may be removed by isochronal annealing.

Radiation embrittlement modelling for reactor pressure vessel steels: I. Brittle fracture toughness prediction

Modelling for the irradiation effect on brittle fracture toughness of reactor pressure vessel (RPV) steel is performed on the basis of the probabilistic model for fracture toughness prediction proposed by the authors earlier. The irradiation effect on parameters controlling plastic deformation and brittle fracture of RPV steels is analyzed. The physical mechanisms are considered which control the cleavage microcrack nucleation for RPV steels in the unirradiated and irradiated states and also in state after post-irradiation annealing. Prediction of the temperature dependence of brittle fracture toughness is performed as applied to irradiated 2.5Cr–Mo–V reactor pressure vessel steel. Modelling of the fluence effect and the phosphorus and copper content effect on brittle fracture toughness is carried out. It is shown that the probabilistic model based on a new formulation for brittle fracture criterion allows the adequate modelling for the irradiation effect on fracture toughness for RPV steel. Application of alternative models is discussed for fracture toughness prediction for irradiated RPV steels.

Thermomechanical Behavior and Modeling Between 350°C and 400°C of Zircaloy-4 Cladding Tubes From an Unirradiated State to High Fluence (0 to 85s˙1024 nm−2,E>1 MeV)

This paper first describes the effect of neutron irradiation on the thermomechanical behavior of stress-relieved Zircaloy-4 fuel tubes that have been analyzed after exposure to five different fluences ranging from nonirradiated material to high burnup. In the second part, a viscoplastic model is proposed to simulate, for different isotherms, 350°C<T<400°C, out-of-flux anisotropic mechanical behavior of the cladding tubes over the fluence range 0<ϕ<100s˙1024 nm−2E>1 MeV. The model, identified for tests conducted at 350°C, has been validated from tests made at 380°C and 400°C. The model is capable of simulating strain hardening under internal pressure followed by a stress relaxation period, the loading producing an interaction between the pellet and cladding. Introduction of a state variable characterizing the damage caused by a bombardment with neutrons into the model has allowed us to simulate the irradiation-induced hardening and creep rate decrease, as well as the saturation noticed after two cycles of irradiation ≅45s˙1024 nm−2E>1 MeV in a pressurized water reactor (PWR). Finally, the numerical simulations show the model is able to reproduce the totality of the thermomechanical experiments. [S0094-4289(00)00202-4]

Superior Charpy impact properties of ODS ferritic steel irradiated in JOYO

The effect of neutron irradiation on Charpy impact properties of an ODS ferritic steel developed by PNC was studied. The miniaturized Charpy V-notch (MCVN) specimens (1.5 × 1.5 × 20 mm) of two orientations (longitudinal, called 1DS-L, and transverse, 1DS-T) were irradiated to fluence levels of (0.3–3.8) × 10 26 n/m 2 ( E n > 0.1 MeV) between 646 and 845 K in JOYO. MCVN specimens before and after the irradiation were subjected to instrumented Charpy impact tests. The test results and fracture surface observations showed that in the unirradiated state the steel showed no ductile-to-brittle transition behavior until 153 K regardless of orientation and the upper shelf energy of the steel was as high as that of a high-strength ferritic steel without dispersed oxide. Such excellent impact properties were essentially maintained after the irradiation although an appreciable decrease in absorbed energy occurred by higher temperature irradiations at and above 793 K.

Small angle neutron scattering investigations of the microstructure of VVER-440-type reactor pressure vessel steel after irradiation at 60°C

The formation of point defects and precipitates after neutron irradiation of VVER-440-type reactor pressure vessel steel was investigated by small angle neutron scattering experiments. Irradiation at 60°C increased the number of point defects, decreased the precipitates, which already exist in the unirradiated state, and formed a new type of fine-scaled precipitates. Post-irradiation annealing near the operational temperature of the nuclear power plants (270°C) provoked a slight decrease of the content of point defects and of the irradiation-induced precipitates. The content of these precipitates which are also present in the unirradiated state did not change by annealing.

Thermocyclic recovery and damage of steel and alloys of various types in un-irradiated and post-irradiated states

Abstract Behaviour of austenitic alloy Fe-0, 3C-20Cr-45Ni-4Mo-B-I, ferritic steel Fe-01C-13Cr-Mo-Ti and titanium alloy Ti-2A1-1, 5Zr during thermocycling in temperature range from 50 to 300 °C have been investigated. These materials have been irradiated in WWR reactor with fluencies 0.44 and 1.2 1021 n/cm2 at T irr ≅ 300 °C. It has been found experimentally that in metals during thermocycling at the temperature not exceeding the irradiation one accelerated recovery of strength and strain properties of irradiated materials and intensive accumulation of thermocyclic damage and as a consequence earlier materials fracture if compared to cycling at constant temperature have been discovered. It has been shown that the thermo-fatigue failure has an intergranular mode with the formation of pores along the grain boundaries. On the basis of obtained experimental data the hypothesis is assumed that two opposite processes take place in metals during thermocycling, they are the recovery of properties and the accumulat...