Effect of cyclotron-injected helium on void-formation characteristics of type 304 stainless steel

Samples of Type 304 stainless steel were injected with helium by cyclotron bombardment to concentrations ranging between $\text{1.1 × 10}{}^{\mathrm{?7}}$ and $\text{1 × 10}{}^{\mathrm{?4}}$ ppma. Following cyclotron injection, the samples were given a variety of heat treatments prior to insertion in EBR-II for irradiation at 450 °C to a total dose of $\text{1 × 10}{}^{21}\text{n/cm}{}^2$. Samples that were not heat treated or that were annealed at 650 °C following cyclotron injection formed few voids and dislocation loops after EBR-II irradiation. This behavior is apparently due to the precipitate clusters that were formed during the helium injection. These precipitates were analyzed by electron microscopic techniques and found to have spherically symmetric strain fields that were of interstitial character. Samples that were annealed at 760 °C following cyclotron injection formed a larger number density of both voids and dislocation loops than did the control sample after EBR-II irradiation. The void volume also exceeded that of the control. Clustering of the dislocation loop population near grain boundaries and precipitate particles was observed in the control and low helium concentration samples.     

Helium embrittlement in type 304 stainless steel

Helium was uniformly injected into small tensile samples of type 304 (austenitic) stainless steel to concentrations of 1 × 10⁻⁷ and 3 × 10⁻⁵ atom fraction helium. Subsequent tensile testing above 540 °C revealed progressive ductility loss—as measured by total elongation at rupture—with increasing temperature. This effect was more severe in samples with the high helium content where elongations at 760 °C were a third or less of those of control samples. Yield and tensile strengths remained unaltered by the presence of helium.Above 650 °C, grain boundary sliding, which results in intergranular cracking, becomes important. The cracks begin as voids on carbide particles which act as obstacles to grain boundary sliding. In the presence of helium, bubbles attach themselves to these carbide particles and serve as void nuclei, thereby accelerating the process. Bubbles were also more prevalent on grain boundaries, dislocations and inclusion particles than isolated in the matrix.     

Neutron irradiation-induced creep of helium pressurized 304L stainless steel capsules

Irradiation-induced creep and swelling have been measured on 1.5 m long pressurized capsules of solution annealed type 304L stainless steel at 385 °C to neutron doses of 45 dpa. The core-midplane results (fixed position) which have a constant average neutron energy and dose rate but varying time are compared to data taken along the length of the capsule which have constant time but varying average neutron energy and dose rates. Additionally, the effect of stress on swelling, the stress dependency of in-reactor creep and the correlation of irradiation-induced creep and swelling are analyzed utilizing the data generated in this experiment. The results of these analyses are then used as a basis for appraising current theories on irradiation creep.     

Carburization of type 304 stainless steel in liquid sodium

Microhardness scans across test specimens were used as a suitable metallurgical technique for determining quantitatively the carburization potential of molten sodium in contact with type 304 stainless steel. The advantage of using hardness data in this manner is that it could be used as a method of analysis of carburization specimens when chemical analyses are not feasible.Results indicated that microhardness values at temperatures of 565, 605, and 650°C could be fitted to curves correlating the hardness to carbon content of liquid sodium. It was found that carbonaceous compounds intentionally mixed in liquid sodium gave a carbon source which produced carburization in the stainless steel test specimens. The capsule methods used for providing and controlling a molten sodium environment for the stainless steel test specimens indicated a need for further study in the use of various capsule-liner materials which are inert to carburization.     

Ferrite formation in neutron-irradiated type 304L stainless steel

Austenitic (γ) to ferrite (α) transformation was observed using transmission electron microscopy in type 304L stainless steel that had been irradiated at ~500°C to fast-neutron (E > 0.1 MeV) fluences greater than ~ 3 × 10²²n/cm². Previous studies on similar unirradiated stainless steels found no such transformation, indicating that the γ → αtransformation was irradiation-induced. The α phase appeared to nucleate on stacking faults, indicating that the presence of large Frank loops was the critical step in the transformation. After an entire grain of austenite had transformed, the only remaining γ phase existed as shells around voids. Coincidence of rapid swelling behavior with γ → α transformation indicated that the two were related, perhaps by reaction of both phenomena to the effects of irradiation and temperature on microchemical segregation. A volume expansion of about 2.5% was found to be associated with the transformation. Inferences are drawn relating this behavior in type 304L steel to the effects of radiation on other reactor structural materials such as type 316 stainless steel, which is also a metastable austenitic composition.     

Radiation-induced segregation in desensitized type 304 austenitic stainless steel

Radiation-induced segregation (RIS) in desensitized type 304 stainless steel (SS) was investigated using a combination of electrochemical potentiokinetic reactivation (EPR) test and atomic force microscopy (AFM). Desensitized type 304 SS was irradiated to 0.43 dpa (displacement per atom) using 4.8 MeV protons at 300 °C. The maximum attack in the EPR test for the irradiated desensitized SS was measured at a depth of 70 μm from the surface. Grain boundaries and twin boundaries got attacked and pit-like features within the grains were observed after the EPR test at the depth of 70 μm. The depth of attack, as measured by AFM, was higher at grain boundaries and pit-like features as compared to twin boundaries. It has been shown that the chromium depletion due to RIS takes place at the carbide-matrix as well as at the carbide-carbide interfaces at grain boundaries. The width of attack at grain boundaries after the EPR test of the irradiated desensitized specimen appeared larger due to the dislodgement of carbides at grain boundaries.     

Stable and unstable crack growth in type 304 stainless steel plate

Experimental and theoretical results on stable as well as unstable fractures for Type 304 stainless steel plates with a central crack subjected to tension force are given.In the experiment using a testing machine with a special spring for high compliance, the transition points from the stable to the unstable crack growth are observed and comparisons are made between the test results and the finite element solutions.A round robin calculation for the elastic-plastic stable crack growth using one of the specimens mentioned above is also given.     

Segregation effects on the corrosion behaviour of a phosphorus-doped AISI type 304L stainless steel

The effect of grain boundary (GB) segregation on intergranular stress corrosion cracking (IGSCC) in hot water environments at 150°C and 250°C was studied in a P-doped AISI type 304L stainless steel. The extent of segregation was measured by an exposure test in boiling 5 N HNO₃ + 8g/L K₂Cr₂O₇ solution as well as by a potentiostatic etch test at +1325 mV (SHE) in 5 N H₂SO₄ solution. Although GB segregation was detected in all the aged specimens, IGSCC was shown by only the specimens aged for 550°C/1000 h. The results suggest that it is the GB chromium depletion, rather than the segregation of phosphorus at the GBs, that controls IGSCC of stainless steels in the hot water environments studied.     

Study on creep-fatigue failure prediction methods for type 304 stainless steel

Creep-fatigue failure is one of the principal failure modes to be avoided in elevated-temperature components of liquid metal fast breeder reactor (LMFBR) plants. To prevent this failure during the plant life with sufficient confidence, accurate and reliable methods should be employed for evaluating creep-fatigue endurance. A number of creep-fatigue tests have been conduced to establish a reliable creep-fatigue design methodology applicable to LMFBR plants in the last two decades but the conditions of these tests are generally far from those expected in actual plants. For the purpose of studying the characteristics of various creep-fatigue life prediction methods in conditions closer to actual plant conditions, the authors initiated creep and creep-fatigue tests for type 304 austenitic stainless steel with a special emphasis on tests with longer durations than past tests. Interim results are summarized in this paper. Two representative life prediction methods, linear damage fraction rule and ductility exhaustion method, were then applied to these test conditions. It was found that both methods can predict the failure lives with reasonable accuracy. Some comparisons were made regarding the characteristics of these two methods.     

Residual stress measurements in irradiated solution-annealed type 304 stainless steel tubing

Immersion density and residual stress measurements were made on solution-annealed type 304 stainless steel capsule tubing irradiated up to fluence levels of 9 × 10²² n/cm² (E > 0.1 MeV). The measured residual stress is dependent on the competition between differential swelling which builds up differential stresses, and irradiation creep which relaxes these stresses. The measurements were analyzed using a bilinear swelling equation formulated with swelling data from the same heat of material. The temperatures and fluence levels of the swelling and slit tube data were each calculated with the same computer code. At high fluence, when swelling was in the steady-state region, the effective irradiation creep rate increased by a factor of about three. Further analysis was made assuming that stress-enhanced swelling and swelling-enhanced irradiation creep were the enhanced relaxation mechanisms.     

In-reactor deformation of solution annealed type 304L stainless steel

Over the past six years at EBR-II, a great deal of information has been obtained on the in-reactor behaviour of solution annealed-Type 304L stainless steel. This information consists of the following: (1) Irradiation induced swelling results in the form of immersion density and transmission electron microscope (TEM) measurements on unstressed material that extends over a temperature range of 395° to 530°C and a neutron fluence range of 1.8 to 9.3 × 10²² n/cm² (E > 0.1 MeV). (2) Irradiation induced creep results from helium pressurized capsules irradiated at a temperature of 415°C. The hoop stress range covered in the experiment was 0 to 27.3 ksi, and the peak neutron fluence obtained to date is 7 × 10²² n/cm² (E > 0.1 MeV). (3) Residual stress measurements (slit tube technique) with complementary TEM gradient studies on stressed and unstressed capsules. (4) Comparative swelling studies of stressed cladding material and unstressed capsule material from encapsulated EBR-II driver fuel experiments over wide ranges of temperature and neutron fluences. The deformation information derived from the four above studies represent an extensive data base from which to obtain an understanding of the in-reactor deformation of austenitic stainless steel. It is the purpose of this paper to review our information on the in-reactor deformation of solution annealed Type 304L stainless steel.     

Thermal fatigue strength of type 304 stainless steel in simulated BWR environment

Thin-walled cylindrical carbon steel specimens were thermally fatigued in a pressurized autoclave. Since high and low temperature pure water were alternately supplied into the autoclave, the specimens were subjected to homogeneous thermal stress through the wall thickness. The thermal fatigue life was defined as the number of cycles to crack penetration to the inside of the cylindrical specimen. The thermal fatigue strength was compared with the mechanical fatigue strength performed in air and in high temperature water. Even if taking account of the Higuchi-Iida formula, which considers the effects of strain rate, dissolved oxygen concentration and water temperature on fatigue life, the thermal fatigue lives of carbon steel were found to be slightly shorter than the mechanical fatigue lives.     

Oxidation of 304 stainless steel in high-temperature steam

An experiment on oxidation of 304 stainless steel was performed in steam between 900°C and 1350°C, using the spare cladding of the reactor of the nuclear-powered ship Mutsu. The temperature range was appropriate for a postulated loss of coolant accident (LOCA) analysis of a LWR. The oxidation kinetics were found to obey the parabolic law during the first period of 8 min. After the first period, the parabolic reaction rate constant decreased in the case of heating temperatures between 1100°C and 1250°C. At 1250°C, especially, a marked decrease was observed in the oxide scale-forming kinetics when the surface treated initially by mechanical polishing and given a residual stress. This enhanced oxidation resistance was attributed to the presence of a chromium-enriched layer which was detected by use of an X-ray microanalyzer. The oxidation kinetics equation obtained for the first 8 min is applicable to the model calculation of a hypothetical LOCA in a LWR, employing 304 stainless steel cladding.     

Corrosion and microstructural aspects of dissimilar joints of titanium and type 304L stainless steel

To link titanium and zirconium metal based (Ti, Zr-2, Ti-5%Ta, Ti-5%Ta-1.8Nb) dissolver vessels containing highly radioactive and concentrated corrosive nitric acid solution to other nuclear fuel reprocessing plant components made of AISI type 304L stainless steel (SS), high integrity and corrosion resistant dissimilar joints between them are necessary. Fusion welding processes produce secondary precipitates which dissolve in nitric acid, and hence solid-state processes are proposed. In this work, various dissimilar joining processes available for producing titanium-304L SS joints with adequate strength, ductility and corrosion resistance for this critical application are highlighted. Developmental efforts made at IGCAR, Kalpakkam are outlined. The possible methods and the microstructural-metallurgical properties of the joints along with corrosion results obtained with three phase (liquid, vapour, condensate) corrosion testing are discussed. Based on the results, dissimilar joint produced by the explosive joining process was adopted for plant application.     

Analytical evaluation method of J-integral in creep-fatigue fracture for type 304 stainless steel

Creep-fatigue crack growth at the operating temperature of LMFBR can be characterized by ΔJ_F and J′ (same as C^*). Type 304 stainless steel, the main structural material of the Japanese LMFBR, shows notable cyclic hardening at elevated temperatures. Evaluation of these J-integrals with the finite-element method is strongly affected by the reference strain range when the cyclic hysteresis' is used as the stress-strain relation.In this paper, an evaluation method for ΔJ_F and J′ with a cyclic stress-strain curve (Δσ – Δ relation) is proposed and verified by experimental results. The evaluation method proposed here does not require cyclic calculations but is monotonic and the effect of the reference strain range is relatively small.     

Radiation damage and surface deformation effects on stainless steel produced by helium-ion bombardment

Radiation damage and surface deformation by neutrons to first-wall CTR materials were simulated by means of 3 MeV helium ions. The irradiation was performed at a CIP cyclotron, with beam intensities of 1–2 μA at room temperature. We have irradiated commercial Romanian, Soviet and Japanese stainless steels (W 4016, 12KH18N10T, W 4541) at doses between 2 × 10¹⁷ and 6 × 10¹⁸ ions per cm². The exfoliations were investigated by means of a TEMSCAN 200-CX electron microscope and a metallographic ORTHOPLAN POL LEITZ microscope. The main post-irradiation characteristics for each type of stainless steel (critical doses for exfoliation, dominant surface morphologies) are discussed. An irradiation facility for obtaining a homogeneous distribution of damage for 27 MeV helium ions (rotating energy degrader) is also presented.