Effects of stress on swelling in 316 stainless steel

Stress was found to increase the magnitude of irradiation-induced swelling in 316 stainless steel. Measurement of the densities of pressurized tube specimens, irradiated at temperatures of ~ 430–475°C to peak fluences of $\text{~ 9 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E > 0.1 MeV}$) in EBR-II, has indicated increased swelling in both the annealed and 20% cold worked conditions of this alloy. Swelling in the annealed specimens was observed to increase linearly with hoop stress up to ~ 20 ksi (130 MPa), whereupon further increases in stress resulted in reduced swelling. Swelling in the cold worked material was linear with stress up to levels of ~ 28 ksi (193 MPa).     

The effect of vacancy loops on the swelling expected for 316 steel under dual-beam irradiation

The theoretical interpretation of simultaneous heavy-ion irradiation and continuous helium injection experiments is extended to include the role of vacancy loops in the evolution of the microstructure. Whilst we find that their inclusion does not alter our previous conclusions concerning the role of the gas in explaining a high-temperature swelling peak, they do influence the details of the swelling response. This detailed influence is such that the results expected from dual-ion simulation experiments should be both more representative of the neutron irradiations and easier to observe than we had previously believed.     

The effect of minor constituents on swelling in stainless steel

The small quantities of solute interstitial elements in stainless steel (C, N and possibly Si) reduce the swelling under neutron irradiation ($\text{～ 2 × 10}{}^{22}\text{neutrons/cm}{}^2$) by more than an order of magnitude between 500 and 600° C over high purity material. The solute interstitials reduce both the numbers and sizes of irradiation-caused voids. Current swelling models ignore — of necessity — this gross effect. Several possible mechanisms are suggested to account for the effect.     

水辐解对反应堆材料SS316腐蚀速率的影响研究

水作为反应堆的主要冷却剂之一,在经过堆芯的辐照区时会产生辐解,生成具有强氧化性的O_2、H_2O_2等产物,这些产物会对材料的腐蚀速率造成影响,进而影响反应堆的活化腐蚀产物源项。在已有理论和模型的基础上,将水辐照分解计算和材料腐蚀速率计算结合起来,以评估水辐照分解对反应堆材料腐蚀速率的影响。根据反应堆的运行工况,计算出冷却回路中水辐解的主要产物O_2和H_2O_2的产额在0.1～10μmol·L~(-1)之间,结合电化学中的混合电位理论,进一步计算得出SS316材料的电化学腐蚀速率在0.012～0.026 g·m~(-2)·h~(-1)范围内。     

Tensile properties of annealed type 316 stainless steel after EBR-II irradiation

The effects of fast reactor neutron irradiation on the tensile properties of annealed Type 316 stainless steel were determined over a wide range of irradiation temperatures and reactor fluences. These effects are described as a function fluence and temperature to $\text{7 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E}{}_{\mathrm{n}}\text{> 0.1}\text{MeV}$) at 430 to 820 °C. The usual flow stress increase and ductility decrease were observed with increasing neutron fluence. Strengthening decreases continuously from 480 °C to ≈ 700 °C with no hardening at or above 760 °C. Elongation values increase with temperature in the 430–540 °C range and generally decrease with temperature above 540 °C.     

The effect of a temperature change on void-swelling in electron-irradiated stainless steel type 316

The void-swelling behaviour of stainless steel type 316 has been investigated in electron-irradiation experiments to doses of 40 dpa, involving a temperature change between 475°C and 575°C, or vice versa, after 20 dpa. In the high/low temperature cycle the swelling was commutative. In the low/high temperature cycle the swelling rate, even at 575°C, was characteristic of 475°C. The dominant point-defect sinks are the dislocations and the effects can be understood in terms of changes in the dislocation density. A low/high temperature cycle is beneficial as long as the dislocations dominate the voids as point-defect sinks and thermal vacancies are unimportant. Implications for the operation of fast breeder reactors are discussed.     

The effect of neutron and He-irradiation on the crack modes of type 316 stainless steel

The objective of this investigation is to determine the crack opening mode (Mode I and Mode II) during in situ HVEM tensile testing and how it is influenced by neutron and helium irradiation, and test temperature. Mode II crack opening was observed as grain boundary sliding initiated by a predominantly Mode I crack steeply intersecting the grain boundary. Mode II crack opening was absent in neutron- or helium-irradiated specimens tested between 400°C and room temperature, but could be restored by a post-irradiation anneal.     

Void swelling and precipitation in a titanium-modified austenitic stainless steel under proton irradiation

The correlation between void swelling and precipitation behavior in a 10% cold worked Fe-16.2Ni-14.6Cr-2.37Mo-1.79Mn-0.53Si-0.24Ti-0.06C alloy was examined with 200 keV proton irradiation. Swelling peak temperature after the proton irradiation to 10 dpa was about 823 K, and void swelling decreased steeply with increase in irradiation temperature from 823 to 923 K. Void swelling increased rapidly from 1.9 to 12.1% with increase in irradiation dose from 20 to 45 dpa at 873 K. Fine intragranular TiC precipitates, which were formed during initial stage of irradiation, dissolved gradually with increase in irradiation dose from 10 to 45 dpa at 873 K, while the amount of precipitation of needle-shaped Fe₂P phase containing titanium increased with increasing dose. The reduction of sink strength of the TiC precipitates due to the dissolution during irradiation was thought to cause the increase of swelling rate with increase in irradiation dose from 20 to 45 dpa at 873 K.     

Quantitative modelling of swelling in M316 steel

The chemical rate theory is widely used to describe the microstructural evolution in a material during irradiation. We describe recent improvements to the theory that increase its predictive capability. The incubation dose prior to the onset of void swelling is modelled by allowing partition of the gas between the various sinks in the microstructure. New dislocation and void sink strengths have been derived incorporating the field effects. These improvements to the theory have been incorporated into a new FACSIMILE computer code designated VS5. The new code has been successfully employed to model void swelling during HVM, VEC and fast reactor irradiation of 316 steel.     

Tritium solubility in SUS-316 stainless steel

Tritium solubility in SUS-316 stainless steel was determined with a gas absorption method, in which tritium gas diluted by protium was used. The tritium absorption experiments were carried out at temperatures of 703, 804 and 903 K under pressures of 10, 30, 50 and 100 torr of tritiated hydrogen gas. The radioactivity of tritium dissolved in the specimen was measured by the method of liquid scintillation counting.The tritium solubility was derived from the experimental data by taking into consideration of isotopic equilibrium among H₂, T₂ and HT molecules. The determined tritium solubility can be expressed by the equation:

     

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.     

Stress corrosion cracking on irradiated 316 stainless steel

Tests on irradiation-assisted stress corrosion cracking (IASCC) were carried out by using cold-worked (CW) 316 stainless steel (SS) in-core flux thimble tubes which were irradiated up to 5×10²⁶ n/m² (E>0.1 MeV) at 310°C in a Japanese PWR. Unirradiated thimble tube was also tested for comparison with irradiated tubes. Mechanical tests such as the tensile, hardness tests and metallographic observations were performed. The susceptibility to SCC was examined by the slow strain rate test (SSRT) under PWR primary water chemistry condition and compositional analysis on the grain boundary segregation was made. Significant changes in the mechanical properties due to irradiation such as a remarkable increase of strength and hardness, and a considerable reduction of elongation were seen. SSRT results revealed that the intergranular fracture ratio (%IGSCC) increased as dissolved hydrogen (DH) increased. In addition, SSRT results in argon gas atmosphere showed a small amount of intergranular cracking. The depletion of Fe, Cr, Mo and the enrichment of Ni and Si were observed in microchemical analyses on the grain boundary.     

The boronizing effect on the radiation shielding and magnetization properties of AISI 316L austenitic stainless steel

The boronizing effect on the radiation shielding properties and magnetization of AISI 316L austenitic stainless steel has been investigated. For this purpose the linear attenuation coefficients of steel have been measured at the photon energies of 662, 1170 and 1332 keV and the results were compared with the calculation at the photon energy of 1-10⁸ keV. It was clearly seen from this work that both the magnetization and radiation shielding properties of the steel have been improved by boronizing process.     

Temperature dependence of the 20% cold worked 316 stainless steel steady state irradiation creep rate

The irradiation creep data from four completed tests have been analysed to show that the steady state irradiation creep rate exhibits a moderate and complex temperature dependence. The irradiation creep tests were performed in the Experimental Breeder Reactor Number II (EBR-II) using beams and pressurized tubes, and in the Oak Ridge Reactor (ORR) and the High Flux Isotope Reactor (HFIR) using pressurized tubes. The data cover the temperature range from 200°C to 585°C, and show that from 200°C to 330°C, the steady state rate increases moderately with increasing temperature. At about 330°C, the steady state rate peaks and rapidly decreases with increasing temperature from 330°C to 370°C. From 370°C to 585°C the steady state rate moderately increases with increasing temperature.     

Dependence of irradiation creep on temperature and atom displacements in 20% cold worked type 316 stainless steel

Irradiation creep studies with pressurized tubes of 20% cold worked Type 316 stainless steel were conducted in the Second Experimental Breeder Reactor. These studies have shown that as atom displacements are extended above 5 dpa and temperatures are increased above 375°C, the irradiation induced creep rate increases with both increasing atom displacements and increasing temperature. The stress exponent for irradiation induced creep remained near unity. Irradiation induced effective creep strains up to 1.8% were observed without specimen failure.     

Helium dilution effect on hydrogen permeation in 316L stainless steel and nickel-base heat-resistant alloys

Effects of inert-gas dilution on hydrogen permeation have been investigated in 316L stainless steel, Inconel 600, Inconel 750, Nimonic 80A and Hastelloy X at 1173 K and 1073 K, by employing a gas-flow system. We used gas mixtures of hydrogen and helium, whose hydrogen concentration ranged from 10^?5 to 10^?1. For the steady-state permeation, the dilution of hydrogen caused no anomalous effects and the permeation rate conformed to Sieverts' law. However, for the transient state, the hydrogen permeation was retarded by the dilution with helium. The retardation effect is discussed in terms of an adsorption model and explained by a decrease in sticking probability at the alloy surface with the dissociative adsorption of hydrogen.