Precision dilatometer analysis of neutron-irradiated nuclear graphite recovery process up to 1673 K

Specimens of two kinds of isotropic nuclear graphite, IG-110U and ETP-10, were neutron-irradiated at fluence of 1.92 × 10²⁴ n/m² (E > 1.0 MeV) at 473 K. The recoveries of the macroscopic lengths of these specimens during isothermal and isochronal annealing at temperatures of up to 1673 K were investigated in a step-wise manner by using a precision dilatometer. The macroscopic lengths after isochronal annealing for 6 h at each temperature decreased gradually as the temperature was increased to 1673 K. The recovery trends of the c-axis and a-axis lattice parameters differed from one another, and from the macroscopic length recovery trends. For the IG-110U specimen, the activation energies (E_a) of macroscopic volume recovery corresponding to annealing at 523–773, 773–923, 923–1073, and 1073–1173 K were found to be 0.15, 0.34, 0.73, and 2.59 eV, respectively. For the ETP-10 specimen, the E_a corresponding to 523–923, 923–1223, and 1223–1373 K were determined to be 0.15, 0.46, and 2.19 eV, respectively. These results indicate that both graphite specimens underwent three or four stages of macroscopic length recovery between 523 K and the annealing temperatures at which their initial lengths were recovered. It is suggested that during the first stage recovery proceeded via the migration of single interstitials along the basal plane and the resulting V-I recombination. In the middle stages, recovery occurred due to the migration of interstitial groups such as C₂ along the basal plane, while in the last stage, it proceeded via through-layer migration of interstitials or migration of single vacancies.     

Dimensional changes and thermal conductivity by annealing and its relation to the defect concentration and stored energy release of neutron-irradiated graphite

The dimensional changes and thermal conductivity with the annealing of fine-grained isotropic graphite IG-110U and ETP-10 irradiated to 0.02 and 0.25 dpa (1.38 x 10²³ and 1.92 x 10²⁴ n/m², E > 1MeV) at a design temperature of <200°C were studied. The irradiated graphite exhibited a small volume expansion and large degradation in thermal conductivity. Post-irradiation annealing experiments were carried out on dimensional changes and thermal conductivity up to 1700°C, and the results were analyzed in terms of changes in the defect concentration of graphite crystals. The rapid recovery of thermal conductivity observed below 200°C in the graphite irradiated to 0.02 dpa was attributed to the annihilation of Frenkel defects, whereas the recovery observed in both dimension and thermal conductivity above 200°C in the graphite irradiated to 0.02 dpa and 0.25 dpa was caused by the annihilation of small interstitial clusters of 4 ± 2 atoms. The role of large clusters of interstitials and vacancies in the changes to smaller dimension than pre-irradiation at high annealing temperatures are discussed. The temperature dependence of stored energy release was estimated from the changes in defect concentration calculated from the recovery of thermal conductivity.     

Investigation of oxidation behaviors of nuclear graphite being developed and IG-110 based on gas analysis

The oxidation behaviors of the nuclear graphite being developed were investigated using gas chromatograph at 873–1373 K. The oxidation experiments were carried out with the gas flow rate of 0.2 L/min and the oxygen concentrations of 7, 10 and 20 mol%. The oxidation reaction began at 973 K and was accelerated with the increase of temperature. At 1173–1273 K, the oxidation was limited by oxygen supplied to graphite and the reaction rate held steady. From 1273 to 1373 K, the oxidation rate increased obviously due to the significant reaction between CO₂ and graphite. At the low temperature regime (973–1073 K), the apparent activation energies with the oxygen mole fractions of 7%, 10% and 20% were 298, 324 and 321 kJ/mol, respectively. Scanning electron microscope was applied to reveal the pore development of the graphite oxidized at different temperatures. The effect of CO combustion at temperature below 1173 K was discussed based on the oxidation behaviors of the graphite being developed and IG-110. It was suggested that the ASTM D7542-15 standard should be adjusted to fit some popular graphite, such as graphite IG-110.     

Cavity nucleation calculations for irradiated metals

A theory of helium-assisted cavity nucleation in irradiated metals is modified and applied to conditions of continuous helium generation. The theory considers the nucleation and growth of cavities by coprecipitation of vacancies, interstitials, and inert gas atoms. Calculations are performed for type 304 stainless steel for comparison with ion irradiation experiments at $\text{~ 2 × 10}{}^{\mathrm{?4}}\text{dpa/s}$, with helium implantation at the rate of ~10^?2 appm/s, to a total damage of ~ 5 dpa, over the temperatures 773–973 K. Total cavity number density calculated ranges from 10²³ m^?3 at 773 K to 10²⁰ m^?3 at 973 K. The calculated incubation time for cavity appearance is 1000–3000 s (0.2–0.6 dpa). The calculated plot of cavity density versus time approximately reproduces the experimental data. Predicted cavity size distributions are roughly bell-shaped, but skewed in favor of larger cavity sizes. Calculated and experimental mean sizes agree within a factor of 3. The predictions of the model are found to change very little when most parameters are varied within reasonable limits. The model is, however, found to be strongly sensitive to cavity : matrix surface energy, as well as the rate that helium atoms are displaced from dislocations.     

Angular Distributions of Resonance Neutron Scattering from Fe-56

Chromium depletion near grain boundaries of austenitic stainless steel during irradiation was investigated. Specimens were kept at 1,473 K for 30 min, and were quenched into the water. Irradiations were done using 400 keV He⁺ ions at 573, 673 and 773 K up to 10dpa with a dose rate of 2.4×10^?4 dpa/s. After irradiation, the Cr concentration profile near the grain boundary was measured using an analytical electron microscope with a 1 nm beam diameter. At 573 K, Cr depletion is small, and its concentration at the grain boundary decreases to 15.5 mass% at 3 dpa from the initial concentration of 18.5 mass%. At 673 and 773 K, Cr concentration at the grain boundary rapidly decreases between 0 and 0.2dpa, and then gradually approaches a constant value, 7.0 mass% at 673 K and 5.0 mass% at 773 K. Two stages are found in radiation induced segregation (RIS) behavior, one stage in which Cr depletion and Ni enrichment balance and another in which Fe depletion and Ni enrichment balance.

These experimental results were compared with the calculations based on the vacancy-induced inverse Kirkendall effect. Predicted Cr segregation at 673 and 773 K above 3dpa agrees with the experimental results. But Cr depletions at low doses which were obtained in the experiments are much faster than calculated. At 573 K in the experiments, depletion is smaller than calculated up to 10dpa.     

CO2 corrosion of IG-110 nuclear graphite studied by gas chromatography

The CO₂ corrosion behavior of IG-110 nuclear graphite has been investigated using the gas chromatography method which allows the continuous analysis of the CO₂/CO gas mixture at the outlet of the corrosion chamber. The effects of temperature and initial CO₂ concentration are studied based on the Arrhenius-type reaction model. From 745 to 995 °C, the Arrhenius curve shows a linear behavior. For higher temperatures, a non-linear behavior is observed. The activation energy is calculated as 210 kJ/mole and is independent of the initial CO₂ inlet concentrations of 10%, 14% and 17%. The corrosion behavior at 1145 °C, in the diffusion-controlled regime, has also been investigated. At this temperature, the interior of IG-110 graphite is severely attacked by CO₂, and the material's surface morphology is changed drastically. A measurement of the corrosion rate against corrosion time shows that the corrosion rate initially increases to a maximum value at a weight loss degree of 30%–35%, after which it begins to decline.     

Resistivity studies of interstitial helium mobility in niobium

The mobility of interstitial helium in Nb and Nb-O alloys was studied in the temperature range of 10–383 K using resistivity measurements. The helium was introduced by radioactive decay of solute tritium (~ 1 at%). At $\text{T < 100 K}$ the resistivity increased due to conversion of tritium trapped at oxygen interstitials to helium. The formation of helium caused a very significant resistance increase at room temperature and above. The results suggest that helium is mobile at temperatures above 295 K and that the precipitation of large helium bubbles occurs along grain boundaries. The mobile helium species may either be single interstitials or small helium clusters. The activation enthalpy for the diffusion of the mobile helium species was estimated to be about 55 kJ/mol (0.66 eV).     

Surface Irradiation Damage of 316 Stainless Steel,Vanadium and Molybdenum by Helium Ion Bombardment

Molybdenum, V and 316 stainless steel were irradiated with 50~150 keV He ions at the temperatures between 413 and 1,298K for total doses ranging 1× 10²²~10ײ³ m^?2, and the characteristics of the surface damage were compared. Severe exfoliation was observed in all of these materials for the irradiation at 413±110 and 748±25K. The number of exfoliated skins was larger than that in literature, and increased nearly in proportion with the total dose. It increased in the order Mo<316SS<. When the dose was low, the amount of erosion increased rapidly with total dose, but tended to be saturated for higher doses than 3×10²² m^?2. It increased in the order Mo<V<316SS at 413±110K, while in the order 316SS<Mo<V at 748±25K. At higher temperatures than 923 K, blisters and porous surface were formed and the exfoliation of skins ceased. The amount of erosion increased with increasing incident ion energy in the energy range between 50 and 150 keV at 413±110K for a total dose of 1×10²² m^?2.     

Measurements of Energy Dependent Doppler Factor for Uranium Dioxide Using a Lead Slowing-Down Time Spectrometer

The energy dependent Doppler factor of resonance absorption in U0₂ was measured in the region of neutron energy between 1 eV and 10 keV using a U0₂ cylindrical sample and a lead slowing-down time spectrometer.

The Doppler factor is herein defined as the ratio of neutron capture rate in the sample at the high temperatures T = 450 and 300° C to that at the low temperature T = 20°C. The Doppler factor measured at T = 450 and 300°C showed the maximum values of 1.225± 0.044 and 1.186±0.040, respectively around the resonance of 66.3 eV. In the energy region above about 200 eV, the Doppler factor decreased with increasing neutron energy. In the energy region below about 70 eV except for region around the resonance of 6.68 eV, the Doppler factor decreased with increasing neutron energy.

The energy dependent Doppler factors were calculated by the collision probability method and the resonance parameters of ENDF/B-IL In the energy region of 1~454 eV, the calculation agreed with the experiment within the error of 5A% which was not unreasonable when one considered the experimental errors of 3.1~4.7%. In the energy region of 454 eV~3 keV, the calculation was systematically 1~6.5% larger than the experiment, though the experimental errors were 2~3.6% in this energy region.     

Tungsten Blister Formation Kinetic as a Function of Fluence,Ion Energy and Grain Orientation Dependence Under Hydrogen Plasma Environment

This work deals with the formation kinetic of tungsten (W) blisters under smooth plasma conditions, i.e. low hydrogen flux and energy in order to analyze the first stages of their formation. In addition, we focus on determining the W grain orientation where blisters grow preferentially. For this purpose, mirror-polished polycrystalline tungsten samples were exposed to hydrogen plasma under fixed hydrogen flux of 2.2 × 10²⁰ m^?2 s^?1, with a fluence in the range of?~?10²⁴ m^?2, ion energy of?~?20, 120 and 220 eV, and sample surface temperature of?~?500 K. The formation of blisters at the surface was investigated using SEM, AFM and EBSD to determine the size, the distribution and the orientation of grain where blisters are formed, respectively. The critical fluence for initiating blisters was established around 2.3 × 10²⁴ m^?2. The evolution of blister size distribution and density is discussed as function of fluence and ion energy. At lower ion energy, i.e. 20 eV, only nanoblisters (less than 150 nm) are observed whatever the fluence value (1.5 and 2.3 × 10²⁴ m^?2). At higher ion energy i.e. 120 and 220 eV, micrometric (~ few to tens of µm) blisters are observed and their density highly depends on fluence. We show that blisters can also be formed on (001) oriented grains contrarily to previous results from the literature where the (111) orientation seemed more favorable. Such information is of importance for tungsten based fusion tokamak operation and design.     

Calculation Program MIVOID for Structure Change of Oxide Fuel during Irradiation

The total cross section of beryllium in the 0.002–0.3 eV neutron energy range was measured by transmission technique at four different temperatures of the sample—300°, 573°, 773° and 973°K.

The scattering kernel was directly calculated from the frequency distribution function, instead of using S(α,β) as practiced in the GASKET-FLANGE method. Integration of the scattering kernel based on the Sinclair model, the Young-Koppel model and the Raubenheimer-Gilat model resulted in little difference between the calculated total cross sections obtained with the three models.

The experimental values are in good agreement with calculations for all measured temperatures of the sample.     

压缩应力下石墨IG-110热膨胀系数的测量及验证

采用应变电测法测量压缩应力状态下石墨IG-110的热膨胀系数,分析不同压缩应力对IG-110热膨胀系数的影响.结果表明,压缩应力对IG-110的热膨胀系数影响显著.与未加载时相比,分别加载20、30、40 MPa压缩应力石墨试样平行加载方向的平均热膨胀系数由3.71×10~(-6) K~(-1)逐渐增大至4.20×10~(-6)、4.41×10~(-6)、4.78×10~(-6) K~(-1),分别提高约13.2%、18.9%和28.8%;而垂直加载方向的平均热膨胀系数则由4.03×10~(-6)K~(-1)逐渐减小至3.80×10~(-6)、3.79×10~(-6)、3.75×10~(-6)K~(-1),分别降低约5.7%、6.0%和6.9%.压缩应力状态下石墨热膨胀系数的变化可能与应力导致石墨内部微裂纹的张开和闭合有关.     

An in Situ Raman Spectroscopy System for Long-term Corrosion Experiments in High Temperature Water up to 673 K

A Raman spectroscopy system has been developed, in order to identify metal oxides formed on the surfaces of metals and steels in high temperature water up to 673 K. A supercritical water loop system including a Raman cell was installed. The design of the loop system is up to 673 K and 40 MPa. The Raman cell has a diamond window without window-to-metal packing. Raman spectrum of alumina plate was measured at room temperature, at 523 K and at 673 K under pressure of 25 MPa. A long-term measurement was also performed at 523 K and 25 MPa for 117.5 h. In all cases intense Raman peaks attributed to alumina were observed. Raman spectrum of anatase particles in suspension was measured at 673 K and 25 MPa. The results show that the Raman spectroscopy system developed in the present study works well not only for plate sample but also for suspension. Raman spectra observed for titanium plate in high temperature water of 673 K and 25 MPa show growth of several Raman peaks with time up to 257 h. The peaks disappeared after cooled down to room temperature. The experimental results have demonstrated importance of in situ Raman spectroscopy.     

Computational Investigation of Nonideal Higher Ionization Carbon Plasma Under Simulated High Heat Fluxes in Fusion Tokamak Reactors

Plasma-facing materials in future large tokamaks will suffer from ablation due to expected hard disruptions, which affects the reactor interior lining tiles and the divertor modules. Ablation and surface evaporation due to the intense heat flux from disruption is associated with ionization of the evolved particulates. Generated ions at such plasma conditions may allow for higher ionization states such that the plasma at the boundary can be composed of electrons, ions (first, second and third ionization) and excited atoms. The boundary layer is dense and tends to be weakly nonideal. The NC State University electrothermal plasma code ETFLOW used to simulate the high heat flux conditions in which the carbon liner tested for simulated heat fluxes for transient discharge period of 100 μs, with FWHM of ~50 μs, to provide a wide range for obtaining reasonable good fits for the scaling laws. Transient events with ~10 MJ/m² energy deposition over short transient of 50–100 μs would produce heat fluxes of 100–200 GW/m². The heat flux range in this simulation is up to 288 GW/m² to explore the generation of carbon plasma up to the third ionization C⁺⁺⁺. The generation of such heat fluxes in the electrothermal plasma source requires discharge currents of up to 250 kA over a 100 μs pulse length with ~50 μs FWHM. The number density of the third ionization is six orders of magnitude less than the first ionization at the lowest heat flux and two orders of magnitude less at the highest heat flux. Plasma temperature varies from 31,600 K (2.722 eV) to 47,500 K (4.092 eV) at the lowest and highest heat fluxes, respectively. The plasma temperature and number density indicate typical high-density weakly nonideal plasma. The evolution of such high-density plasma particles into the reactor vacuum chamber will spread into the vessel and nucleate on the other interior components. The lifetime of the PFCs will shorten if the number of hard disruptions at such extreme heat fluxes would be increasing, resulting in major deterioration of the armor tiles.     

Effects of pre-irradiation treatments and minor composition modification on neutron irradiation embrittlement of 316 stainless steel

The effects of the combination of heavy cold work and low temperature aging (873–1073 K, 100 h), and minor composition modification on the irradiation embrittlement of 316 stainless steel were investigated. The samples were irradiated by JMTR at JAERI to the dose level of 2.5 × 10²⁴ n/m²$\text{(E > 1 MeV)}$ at 823 and 923 K and tensile tested between R.T. and 1023 K. The embrittlement was compared from the standpoint of ductility survival ratio. The lowering of carbon content caused severer high temperature helium embrittlement in the solution treated condition. The heavy cold work and low temperature aging treatments could not improve the high temperature embrittlement compared with the solution treated condition. Titanium addition was beneficial especially for the reduction of the irradiation temperature sensitivity to the high temperature ductility.     

Oxidation of Iron at Oxygen Partial Pressures of 1, 104.3 and 105Pa in Temperature Range from 523 to 673 K

Abstract

The oxidation of iron was studied by means of Rutherford backscattering spectroscopy in the temperature range from 523 to 673 K under the oxygen partial pressures of 1, 10^4.3 and 10⁵ Pa for up to 7 d. It was found that the ionic diffusion in oxides is the rate-determining step of oxidation, since the oxidation of iron obeyed the parabolic rate law in the temperature and oxygen partial pressure ranges studied. As was observed at 573 K⁽¹⁾, two-stage oxidations were observed at temperatures lower than around 650 K, and one-stage oxidations at higher temperatures. The activation energies for the oxidation in the second stage were obtained to be 133±10, 149±17 and 144±6 kj.mol^?1 at 1, 10^4.3 and 10⁵ Pa, respectively. The activation energies for oxidation obtained are nearly constant regardless of the oxygen partial pressure, but the absolute values of the parabolic rate constants are determined by the fraction of the coverage on magnetite with hematite. The changes in the temperature dependence of the parabolic rate constants were found around 670 K at 10^4.3Pa and around 523 K at 10⁵Pa, where the oxidation mechanism may change from the simultaneous growth of magnetite and hematite at higher temperatures to the dense hematite formation at lower temperatures. Comparing the temperature dependence of the corrosion rate in water with that of the oxidation rate in gas phase, the corrosion rate in water becomes considerably larger than the oxidation rate in gas phase below 523 K, where the metal dissolution may be dominant in the corrosion process in water.     

Dimensional Changes in Irradiated (Th,U)O2

Effects of irradiation on the dimension and microstructure in (Th,U)O₂ pellets were examined by measurements of lattice parameter and bulk density changes, and observations of pore structures. The concentrations of fission-induced defects and the damage volume were estimated by a simple model. Both macroscopic and microscopic dimensional changes were found to increase initially with fission dose and then fall off. The difference between macroscopic and microscopic ingrowths increased with dose, suggesting that fission-induced interstitials would cluster or go to sinks and the concentration of vacancies would be in excess of that of interstititials. The damage volume for vacancies was estimated to be about 1x10^?22m³·fiss.^?1, and almost agreed with that for fission Xe release. Observations of the pore structure indicated that the volume fraction of pores smaller than 2–3 μm decreases with irradiation and the distribution of pore size shifts toward the larger side.     

Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

Dissolution behavior of (U,Zr)O2-based simulated fuel debris in nitric acid

To explore the possibility of dissolving fuel debris into nitric acid as a potential pre-treatment for waste treatment in which the U and Pu are removed from the inventory, dissolution tests of U_1?xZr_xO₂ and (U,Pu)_1?xZr_xO₂ were carried out in 6 M HNO₃ at 353 K. At the end of the dissolution test (after 4 h), the ratio of dissolved uranium decreased with an increase in the Zr contents, x. While the dissolution of U-rich samples was congruent, a preferential leaching of U was observed with Zr-rich samples. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). The IDR decreased from 10^?5 down to 10^?10 mol cm^?2 min^?1 as x increased from 0 to 0.95. From these findings, dissolution with HNO₃ is expected to be only applicable in U-rich part of fuel debris (x < 0.3) if the dissolution in 6 M HNO₃ at 353 K is assumed. Application of complexing acids, such as mixture of HNO₃ and HF, should be considered to increase the dissolution rate of the Zr-rich part.     

Oxidation behavior of IG and NBG nuclear graphites

This work studies the oxidation-induced characteristics of four nuclear graphites (NBG-17, NBG-25, IG-110, and IG-430). The oxidation characteristics of the nuclear graphites were measured at 600 °C. The surface properties of the oxidation graphites were characterized by means of scanning electron microscopy, X-ray photoelectron spectroscopy, and contact angle methods. The N₂/77 K adsorption isotherm characteristics, including the specific surface area and micropore volume, were investigated by means of BET and t-plot methods. The experimental results show an increase in the average pore size of graphites; they also show that oxidation produces the surface functional groups on the graphite surfaces. The surface area of each graphite behaves in a unique manner. For example the surface area of NBG-17 increases slightly whereas the surface area of IG-110 increases significantly. This result confirms that the original surface state of each graphite is unique.