Comparison of TSSD results obtained by differential scanning calorimetry and neutron diffraction

Due to their low absorption cross-section for neutrons, Zr alloys are used for reactor core components. The terminal solid solubility (TSS) for hydrogen in these alloys is very low - in Zr-2.5 wt% Nb, used to fabricate pressure tubes for CANDU (CANDU-CANada Deuterium Uranium is a registered trademark of Atomic Energy of Canada Ltd.) power reactors, the TSS is ∼0.7 at.% H at 300 °C. The mechanical properties of the components may deteriorate when their hydrogen concentration exceeds TSS. Therefore, accurate values of the TSS are needed to assess the operating and end-of-life behaviours of these components. Differential scanning calorimetry (DSC) is used to measure the TSS of hydrogen in Zr alloys. Three distinct features are marked on a typical DSC heat flow curve when the material is being heated and the hydrides are dissolving; ‘peak temperature’, ‘maximum slope temperature’ and ‘completion temperature’. Usually, the maximum slope temperature, being about the average of the three temperatures, is interpreted as the TSS temperature for hydride dissolution (T_TSSD). A set of coordinated DSC and neutron diffraction measurements have been carried out to identify the features of the heat flow signal that closely correspond to the T_TSSD. Neutron diffraction was chosen because hydrides generate distinctive diffraction peaks whose intensities approach zero at the transition temperature - an unambiguous indication of dissolution. Neutron diffraction shows that the temperature of hydride dissolution correlates closely with the DSC peak temperature.     

Influence of metallurgical variables on delayed hydride cracking in Zr-Nb pressure tubes

During service, Zr-2.5Nb pressure tubes of nuclear power reactors may be prone to suffer from crack growth by delayed hydride cracking (DHC). For a given hydrogen plus deuterium concentration there is a critical temperature (T_C) below which DHC may occur. In this work, T_C was measured for specimens cut from pressure tubes made in Canada (CANDU) and in Russia (RBMK). Hydrogen was added to the specimens to get concentrations ranging from 24 to 60 wt ppm. It was found that T_C was higher than the corresponding precipitation temperature. The crack propagation velocity (V_P), measured in axial direction, increases from a minimum at T_C to a maximum at a temperature close but higher than the precipitation temperature. At lower temperatures, when hydride precipitates are present in the bulk, V_P follows an Arrhenius law: V_P = A exp(−Q/RT), with an activation energy Q of 66-68 kJ/mol for both tubes. The RBMK material presented lower velocities than CANDU one.     

Delayed hydride cracking in Zr-2.5Nb tube with the cooling rate and the notch tip shape

The objective of this study is to demonstrate the feasibility of the Kim’s delayed hydride cracking (DHC) model. To this end, this study has investigated the velocity and incubation time of delayed hydride cracking (DHC) for the water-quenched and furnace-cooled Zr-2.5Nb tubes with a different radius of notch tip. DHC tests were carried out at constant K_I of 20 MPa √m on cantilever beam (CB) specimens subjected to furnace cooling or water quenching after electrolytic charging with hydrogen. An acoustic emission sensor was used to detect the incubation time taken before the start of DHC. The shape of the notch tip changed from fatigue cracks to smooth cracks with its tip radius ranging from 0.1 to 0.15 mm. The DHC incubation time increased remarkably with the increased radius of the notch tip, which appeared more strikingly on the furnace-cooled CB specimens than on the water-quenched ones. However, both furnace-cooled and water-quenched CB specimens indicated little change in DHC velocity with the radius of the notch tip unless their notch tip exceeded 0.125 mm. These results demonstrate that the nucleation rate of hydrides at the notch tip determines the incubation time and the DHC velocity becomes constant after the concentration of hydrogen at the notch tip reaches terminal solid solubility for dissolution (TSSD), which agrees well with the Kim’s DHC model. A difference in the incubation time and the DHC velocity between the furnace-cooled and water-quenched specimens is attributed to the nucleation rate of reoriented hydrides at the notch tip and the resulting concentration gradient of hydrogen between the notch tip and the bulk region.     

The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behavior

Differential scanning calorimetry (DSC) has been applied to elucidate the terminal solid solubility (TSS) of hydrogen in Zircaloy-2 cladding tubes and spacer bands irradiated in commercial BWRs. While recovery of irradiation defects during the first heating stage of as-irradiated specimens made the DSC peak of hydride dissolution dull or broader, no significant difference was detected in the TSS between unirradiated and irradiated Zircaloy-2, irrespective of fast neutron fluence. The effect of post-irradiation annealing on TSS was also examined. The results suggest almost no interaction between irradiation defects and dissolved hydrogen or hydrides at temperatures around 300 °C. Using the present TSS data and reported hydrogen- and hydride-related properties, a microscopic analysis code HYMAC for analyzing hydride behavior in cladding tube with textured grains was constructed. Stress-induced preferential precipitation and dissolution of hydrides were reproduced by adopting a TSS sub-model in which the solubilities decrease in proportion to stress normal to the habit plane in grains and to grain faces. Analyzed results by the code were consistent with typical experimental results of hydride behavior.     

Comment on ‘The first step for delayed hydride cracking in zirconium alloys’ by G.A. McRae et al., J. Nucl. Mater. 396 (2010) 130-143

The aim of this paper is a reply to McRae et al.’s paper entitled “The first step for delayed hydride cracking (DHC) in zirconium alloys” claiming that the first step of DHC is hydrogen diffusion, not nucleation of hydrides as demonstrated by Kim’s new model. Despite the authors’ claim that the crack tip concentration is higher than the bulk concentration due to the stress gradient, their claim violates the thermodynamic principle that the stressed region should have a lower potential of hydrogen or lower hydrogen solubility than the unstressed region. Furthermore, it is demonstrated that the Diffusion First Model (DFM) proposed by the author is defective in terms of kinetics because hydrogen diffusion before hydride nucleation just governs the rate of hydride nucleation, neither the rate of hydride growth nor the crack growth rate (CGR).     

Irradiation and annealing effects on delamination toughness in carbon/epoxy composite

Gamma irradiation to various doses (4.8-27.2 MGy) was performed on unidirectional carbon fiber/epoxy resin composite plates. Unidirectional composite coupons irradiated to various doses were annealed at 180 and 250 °C, in vacuum. The strain energy release rate G_IC, as a measure of delamination fracture toughness, was determined by Mode I fracture testing on double cantilever beam coupons. The glass transition temperature (T_g) of the tested coupons matrices was determined in DMA tests. The effects of irradiation and annealing on G_IC values - the mean values of 10 propagation points (G_IC,mean) and that of fracture initiation (G_IC,init) - were established. These values were analyzed as a function of irradiation dose and annealing temperatures, having in mind glass transition temperature values changes, as well as the possible mechanisms and phenomena of irradiation and annealing.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Collision cascade temperature

Interaction of a projectile with a solid has been considered in detail. It has been found that any collision cascade generated by a projectile can be characterized by the average kinetic energy of cascade atoms that represents an “instantaneous temperature” of the cascade during its very short lifetime (10⁻¹² s). We refer to this value as the “dynamic temperature” in order to emphasize the fact that cascade atoms are in a dynamic equilibrium and have a definite energy distribution. The dynamic temperature defines the electron distribution in the cascade area and, hence, the ionization probability of sputtered atoms. The energy distribution of cascade atoms and, as a consequence, the dynamic temperature can be found experimentally by measuring the energy distribution of sputtered atoms. The calculated dynamic temperature has been found to be in good agreement with the experimental data on ion formation in the case of cesium and oxygen ion sputtering of silicon. Based on the developed model we suggest an experimental technique for a radical improvement of the existing cascade sputtering models.     

Microstructural degradation mechanisms during creep in strength enhanced high Cr ferritic steels and their evaluation by hardness measurement

There are two creep regions with different creep characteristics: short-term creep region “H”, where precipitates and subgrains are thermally stable, and long-term creep region “L”, where thermal coarsening of precipitates and subgrains appear. In region “H”, the normalized subgrain size (λ-λ₀)/(λ^∗-λ₀) has a linear relation with creep strain and its slope is 10ε⁻¹. But, region L is the time range in which the static recovery and the strain-induced recovery progress simultaneously. In this region, the static recovery accelerates the strain-induced recovery, and subgrain size is larger than that line which neglects the contribution of the static recovery. In region “L”, the Δλ/Δλ^∗-strain present a linear relation with a slope 35ε⁻¹. There is a linear relation between hardness and subgrain size. Hardness drop, H₀ − H, as a function of Larson-Miller parameter can be a good measure method for assessment of hardness drop and consequently degradation of microstructure. Hardness drop shows an identical slope in creep region “H”, whereas hardness drop due to thermal aging and creep in region “L” show together a similar slope. In region “H”, degradation of microstructure is mainly due to recovery of subgrains controlled by creep plastic deformation, and precipitates do not have a major role. However, in creep region “L”, there are three degradation mechanisms that accelerate creep failure; (1) strain-induced recovery of subgrains due to creep plastic deformation, (2) static-recovery of subgrains and precipitates and (3) strain-induced coarsening of precipitates due to the appearance of static-recovery.     

Heat capacity of hydrogenated Zircaloy-2 and high Fe Zircaloy

Heat capacities (C_p) of non-hydrogenated and hydrogenated Zircaloy-2 and high Fe Zircaloy were measured in the temperature range from 350 to 873 K, using a differential scanning calorimeter. The hydrogen concentrations in the two types of alloys ranged from 26 to 1004 ppm. The C_p values of the as-received alloys with 26-29 ppm hydrogen were in good agreement with literature data for low hydrogen Zircaloys. From this finding and observation of almost the same enthalpy changes for hydride dissolution for both alloys, it was concluded that there was no difference in C_p values between the two types of hydrogenated Zircaloys. The dissolution enthalpy of hydrides calculated from C_p data was 41.0 kJ/g-atom H. For Zircaloy-2 samples with higher hydrogen concentrations than 700 ppm, the phase transition from α+δ to α+β was observed at the eutectoid temperature of 824-827 K. Two types of models describing an additional heat capacity due to the hydride dissolution were presented based on the present C_p data and previously derived terminal solid solubility of hydrogen.     

The first step for delayed hydride cracking in zirconium alloys

Two models for delayed hydride cracking (DHC) in zirconium alloys are distinguished by their first step:

-

The loading of a crack induces hydride precipitation. The hydride is postulated to create a hydrogen concentration gradient, where the bulk concentration is greater than that at the crack tip. This concentration gradient is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the precipitate first model (PFM).

-

The tensile stress at the crack tip induces a gradient in chemical potential that promotes the diffusion of hydrogen to the crack tip. Hydrides form if the hydrogen concentration reaches the solubility limit for hydride precipitation. The mechanism is postulated to create a hydrogen concentration gradient, where the bulk concentration is lower than that at the crack tip. The gradient in chemical potential is taken as the driving force for diffusion of hydrogen to the crack tip, and subsequent hydride growth. This model is called the diffusion first model (DFM).

The second model, DFM, is developed. This model is shown to describe the main features of the experimental observations of DHC, without invoking new phenomena, such as reduction in the solubility limit for precipitation of hydride, as required by the PFM.     

Nuclear resonant reflectivity with standing waves for the investigation of a thin Fe layer buried inside a superconducting Si/[Mo/Si]45/Fe/Nb multilayer

A special multilayer sample Si/[Mo/Si]₄₅/⁵⁷Fe/Nb has been prepared for the depth selective investigations of the hyperfine fields in thin iron layer at low temperatures above and below the superconducting transition in the top Nb layer (T_c ∼ 8 K) by means of the nuclear resonant reflectivity with standing waves. The periodic multilayer [Mo/Si]₄₅ below the iron layer in our sample was used as “a standing wave generator”. A weak magnetic hyperfine splitting in the ⁵⁷Fe layer was detected just at low temperature. A slight variation of the nuclear resonant reflectivity time spectra measured above and below T_c was observed. At first it was supposed that this change of the spectrum shape is caused by the spatial modulation of ferromagnetic domains in the ⁵⁷Fe layer caused by a proximity effect. A closer analysis, however, reveals that the spectrum variations are due to just the changes of the relative weights of the magnetic and paramagnetic phases in ⁵⁷Fe layer.     

The effects of Co γ-ray on poly(ethylene-co-vinyl acetate)/carbon black composites

Cables used in a nuclear power plant are irradiation suppressing ones. Until now, researches on the irradiation suppressing cables have mainly been focused on insulation materials. Therefore, in this paper, the non-isothermal crystallization behaviors and degradation characteristics of ethylene vinyl acetate-carbon black (EVA-CB), used as a shielding material, were investigated by means of the Differential scanning calorimetry (DSC) and chemiluminescence analyzer (CL). The specimens were cooled after removing thermal history at 150 °C for 5 min by changing the cooling rates to 5, 7.5, 10, 15 and 20 °C/min with DSC. In addition, after maintaining a thermal equilibrium at each temperature of 25, 50, 75, 100, 125, 150 and 175 °C, their thermoluminescence was measured for 20 min with CL equipment. The ⁶⁰Co γ-ray was used for irradiation. T_c, T₀, T_∞ and t_1/2 in the DSC experiments are found to decrease gradually as radiation dose increases. Secondly, with the CL experiment, the 0.1, 0.25 and 0.5 MGy EVA-CB composites were found to show a much smaller thermoluminescence than the intact EVA-CB composites, while the 0.75 and 1 MGy EVA-CB composites were found to show a much higher thermoluminescence than ones.     

Oxidation kinetics and oxygen diffusion in low-tin Zircaloy-4 up to 1523 K

This paper deals with the study of oxidation kinetics and the identification of oxygen diffusion coefficients of low-tin Zy-4 alloy at intermediate (973 K ? T ? 1123 K) and high temperatures (T ? 1373 K). Two different cases were considered: dissolution of a pre-existing oxide layer in the temperature range 973 K ? T ? 1123 K and oxidation at T ? 1373 K. The results are the following ones: in the temperature range 973-1123 K, the oxygen diffusion coefficient in α_Zr phase can be expressed as D_α = 6.798 exp(−217.99 kJ/RT) cm²/s. In the temperature range 1373-1523 K, the oxygen diffusion coefficients in α_Zr, β_Zr and ZrO₂, were determined using an ‘inverse identification method’ from experimental high temperature oxidation data (i.e., ZrO₂, and α_Zr(O) layer thickness measurements); they can be expressed as follows: D_α = 1.543 exp(−201.55 kJ/ RT) cm²/s, D_β = 0.0068 exp(−102.62 kJ/ RT) cm²/s and D_ZrO2=0.115exp(−143.64kJ/RT)cm²/s. Finally an oxygen diffusion coefficient in α_Zr in the temperature range 973 K ? T ? 1523 K was determined, by combining the whole set of results: D_α = 4.604exp(−214.44 kJ/RT) cm²/s. In order to check these calculated diffusion coefficients, oxygen concentration profiles were determined by Electron Probe MicroAnalysis (EPMA) in pre-oxidized low-tin Zy4 alloys annealed under vacuum at three different temperatures 973, 1073 and 1123 K for different times, and compared to the calculated profiles. At last, in the framework of this study, it appeared also necessary to reassess the Zr-O binary phase diagram in order to take into account the existence of a composition range in the two zirconia phases, α_ZrO2 and β_ZrO2.     

Evaluation of ductile–brittle transition temperature before and after neutron irradiation for RPV steels using small punch tests

Small punch (SP) tests were performed to evaluate the ductile–brittle transition temperature before and after a neutron irradiation of reactor pressure vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the conventional Charpy tests and the Master Curve fracture toughness tests in accordance with the American Society for Testing and Materials (ASTM) standard E1921. Small punch specimens were taken from a 1/4t location of the vessel thickness and machined into a 10 mm × 10 mm × 0.5 mm dimension. The specimens were irradiated in the research reactors at Korea Atomic Energy Research Institute Nuclear Research Institute in the Czech Republic at the different fluence levels of about 290 °C. Small punch tests were performed in the temperature range of RT to −196 °C using a 2.4 mm diameter ball. For the materials before and after irradiation, the small punch transition temperatures (T_SP), which are determined at the middle of the upper small punch energies, showed a linear correlation with the Charpy index temperature, T_41 J. T_SP from the irradiated samples was increased with the fluence levels and was well within the deviation range of the unirradiated data. However, the transition temperature shift from the Charpy test (ΔT_41 J) shows a better correlation with the transition temperature shift (ΔT_SP(E)) when a specific small punch energy level rather than the middle energy level of the small punch curve is used to determine the transition temperature. T_SP also had a correlation with the reference temperature (T₀) from the Master Curve method using a pre-cracked Charpy V-notched (PCVN) specimen.     

“Water window” compact, table-top laser plasma soft X-ray sources based on a gas puff target

We have developed compact, high repetition, table-top soft-X-ray sources, based on a gas puff target, emitting in “water window” spectral range at λ = 2.88 nm from nitrogen gas target or, in 2-4 nm range of wavelengths, from argon gas target. Double stream gas puff target was pumped optically by commercial Nd:YAG laser, energy 0.74 J, pulse time duration 4 ns. Spatial distribution of laser-produced plasma was imaged using a pinhole camera. Using transmission grating spectrometer, argon and nitrogen emission spectra were obtained, showing strong emission in the “water window” spectral range. Using AXUV100 detector the flux measurements of the soft-X-ray pulses were carried out and are presented.These debris free sources are table-top alternative for free electron lasers and synchrotron installations. They can be successfully employed in microscopy, spectroscopy and metrology experiments among others.     

Threshold stress intensity factor for delayed hydride cracking of a recrystallized N18 alloy plate along the rolling direction

The objective of this study is to obtain the threshold stress intensity factor, K_IH, for an initiation of delayed hydride cracking in a recrystallized N18 (Zr-Sn-Nb-Fe-Cr) alloy plate which was manufactured in China, gaseously charged with 60 ppm of hydrogen by weight. By using both the load increasing method and load drop method, the K_IH’s along the rolling direction were investigated over a temperature range of 150-255 °C. The results showed that K_IH along the rolling direction was found to be higher in the load increasing method than that in the load drop method. In the load increasing method, K_IH’s of the N18 alloy plate appeared to be in the range of and K_IH in the load drop method appeared to be in the range of . This means that the N18 alloy plate has high tolerance for DHC initiation along the rolling direction. The texture of a N18 alloy plate was investigated using an X-ray diffraction and the K_IH was discussed based on texture and analytically as a function of the tilting angle of hydride habit planes to the cracking plane.     

Properties of FINEMET with Fe partially replaced by Mn

Thin films and foils of Fe_73.5−xSi_13.5B₉Cu₁Nb₃Mn_x, the FINEMET based amorphous and nanocrytalline alloys with high Mn doping (x = 9, 11, 13, 15 at%), were studied by means of transmission electron microscopy (TEM), X-ray diffraction (XRD), differential scanning calorimetry (DSC) and with Mössbauer spectroscopy (MS), as-quenched (a-q) and after annealing (a). Mn, partially replacing Fe, causes common crystallisation temperature T_cr for the identified crystal structures, decrease of the lattice constants a₀, c₀, decrease of hyperfine parameters: magnetic field H_hf and isomer shift IS for amorphous phases and in consequence the Curie temperature T_Cu.     

Delayed Hydride Crack Velocity of Zirconium Alloys with the Direction of an Approach to Temperature

The delayed hydride cracking (DHC) tests were conducted on Zr-2.5Nb compact tension specimens with the test temperatures reached by a heating and a cooling. The Zr-2.5Nb specimens were either furnace-cooled or water-quenched after a hydrogen charging treatment to contain 10 to 100 ppm H. On an approach to the test temperatures by a cooling, both the Zr-2.5Nb specimens showed the DHC velocity increasing with an increasing temperature over a temperature range of 100-300°C, irrespective of the cooling rate. However, on an approach to the test temperatures by heating, the furnace-cooled Zr-2.5Nb showed a DHC arrest at temperatures over 180°C and no DHC at 250°C, and the water-quenched ones did have a DHC growth, even at 250°C. Using Kim's DHC model we elucidate the DHC arrest in the furnace-cooled Zr-2.5Nb at temperatures over 180°C and the DHC growth in the water-quenched specimen, even at 250°C, upon an approach by a heating.     

Hydride behavior in Zircaloy cladding tube during high-temperature transients

In order to study the hydride behavior in high burnup fuel cladding during temperature transients expected in anticipated operational occurrences and accidents, unirradiated hydrided Zircaloy-4 cladding tubes were rapidly heated to temperatures ranging from 673 to 1173 K and annealed for holding time ranging from 0 to 3600 s. Hydrides were localized in the peripheral region of the cladding tubes prior to the annealing, as observed in high burnup fuel cladding. The localized hydride layer (hydride rim) was annealed out, and the radial hydride distribution became uniform after the annealing at 873 K for 600 s, 973 K for 60 s, or 1173 K for 0 s. The annealing out of the hydride rim is caused by the phase transformation from the α + δ phase to the α + β or β phase in the hydride rim and the subsequent drastic increase in the solid solubility and diffusion of hydrogen in Zircaloy. On the other hand, the radial distribution and morphology of hydrides did not change at lower temperatures: Thus, the hydride remains almost intact below the phase transformation temperature for the short time ranges.