The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.     

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.     

Swelling and radiation creep of austenitic corrosion-resistant steel irradiated by neutrons in wide dose and temperature ranges

The results of a study of the swelling and in-reactor creep of EI-847, EP-172, and ChS-68 austenitic steel after irradiation in materials science assemblies in the range 330–700°C and damaging dose 20–96 dpa are presented. The temperature dependences of the volume change of steel were obtained from measurements of the diameter of unloaded ampuls. It is shown that the swelling of the steel increases linearly with increasing tangential stress. The modulus of in-reactor creep in the interval 410–630°C for the steel investigated in the cold-deformed state varies in the range (0.5–3)·10^–6 MPa^–1·dpa^–1. For lower and higher temperatures, the creep modulus increases to (5–8)·10^–6 MPa^–1·dpa^–1.     

In-reactor tubular creep of Zircaloy-2 at 260 to 300 °C

Tubular specimens of Zircaloy-2, 23 mm diameter, have been creep tested in-reactor at 260 to 300°C (530 to 570 K). The specimens were biaxially stressed by internal pressure, with transverse stresses from 100 to 300 MN/m². Zircaloy-2 was tested in three conditions; 20% cold drawn, 70% tube reduced then stress-relieved and annealed.All creep curves, both in and out of neutron fluxes, can be represented by straight lines on log strain-log time curves. Fast neutron flux increased the slopes of the log-log creep curves of the cold-worked materials. These slopes increased from 0.24–0.27 for unirradiated specimens (and specimens in the thermal neutron flux) to 0.42–0.47 for specimens in a fast neutron flux. This means that creep rate does not diminish with time as rapidly in-reactor as out-reactor. The creep behaviour of the annealed Zircaloy-2 was little affected by fast neutron flux.     

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.     

In-reactor investigations of the creep of structural materials

Conclusions 1. A series of in-reactor tests was performed on a sample used to study radiation creep in 00X16H15M3B steel, XHM1 chrome-nickel alloy, the zirconium based alloys é110 and é635, and the vanadium-based alloy BTX8. The radiation creep modulus (in units of Pa⁻¹·(displacements/atom)⁻¹ equals 1.7·10⁻¹¹ for 00X16H15M3B steel, 4.6·10⁻¹¹ for XHM alloy with fluence up to 2.3·10²⁰ cm⁻² and 1.6·10⁻¹¹ for a fluence above 1·10²¹ cm⁻², (4.6–4.9)·10⁻¹¹ for é110 alloy, and 1.8·10⁻¹¹ for é635 alloy. For the alloy BTX8, at stresses below half the yield point and t=450°C, the modulus equals 3.3·10⁻¹² Pa⁻¹·(displacements/atom)⁻¹. At a higher stress, the deformation rate of the alloy increases progressively. 2. In the investigation of the temperature dependence of in-reactor creep of the alloy é110, it was found that at 350–370°C and higher, the thermal creep makes the predominant contribution to deformation. In the experimental range 370–455°C, the thermal activation energy of in-reactor creep was determined to be 36 ± 8 kcal/(g·atom). At temperatures below 350°C the creep of the alloy é110 is a temperature-independent radiation-stimulated process. 3. In the case of tests of zirconium alloys, a previously unobserved phenomenon of periodic rapid deformation of the material against the background of creep at stresses even well below the yield point of the irradiated material was discovered. The effect was manifested at a temperature of about 230°C. As the temperature increases up to 290°C and higher, no plastic movements are observed. Translated from Atomnaya énergiya, Vol. 80, No. 5, pp. 386–391, May, 1996.     

Irradiation-creep in a materials testing reactor

Helical springs made from titanium, zirconium, Nimonic PE 16 alloy and two austenitic stainless steels namely Firth-Vickers FV 548 and a steel with a composition within the AISI 316 specification have been irradiated in the Dounreay Materials Testing Reactor, at temperatures between 85 °C and 100 °C and to neutron fluences up to about 4 × 10²⁴ m^?2 (> 1 MeV), whilst loaded in tension. Irradiation-creep was observed in all the materials studied and initial strain rates/unit stressneutron ranged from 1 to 1.9 × 10^?35 m⁴/N for Nimonic PE 16 and zirconium respectively. Data obtained from an earlier experiment are re-presented and compared with the present results.Springs which received no heat-treatment after coiling unwound during irradiation at rates which were independent of the supported loads. The phenomenon is attributed to the relaxation of internal stresses (produced during the manufacture of the springs) with an irradiation-creep constant which is an increasing function of prior cold-work.A mechanism of irradiation-creep is proposed which involves the re-arrangement of the dislocation network in a crystal as the dislocations climb by absorbing interstitials produced by irradiation.     

Correlation of neutron and heavy-ion damage: I. The influence of dose rate and injected helium on swelling in pure nickel

Displacement damage structures in pure nickel at the 1 dpa level are compared for two widely disparate damage rates, 10⁻⁷ dpa/s for neutron irradiations and 3 X 10⁻³ dpa/s for self-ion bombardments over a range of temperatures spanning those for void formation. Peak swelling at about 0.7% is found at 400° and 600°C, respectively. At equivalent swelling temperatures, voids in the ion-bombarded material are larger and fewer than those from neutron irradiation, especially at temperatures above the peak swelling temperature.Additions of 20 appm He, matching that generated in the neutron irradiations, were made to the ion-bombarded nickel either prior to ion bombardment (preinjection) or during ion bombardment (simultaneous injection). This helium caused increased swelling at the upper and lower temperature extremes. Simultaneously implanted helium did not otherwise significantly affect microstructures, whereas preinjected helium increased the dislocation density and caused more but smaller voids over the full temperature range of swelling.     

In-reactor stress relaxation of selected metals and alloys at low temperatures

Sttess relaxation of bent beam specimens under fast neutron irradiation at 340 and 570 K has been studied for a range of materials, as follows: several stainless steels, a maraged steel, AISI4140, Ni, Inconel X-750, Ti, Zircaloy-2, Zr-2.5% Nb and Zr₃ Al. All specimens were in the annealed or solution-treated condition. Where comparisons were possible, the creep coefficients derived from the stress relaxation tests were found to be consistent with other studies of irradiation-induced creep. The steels showed the lowest rates of stress relaxation; the largest rates were observed with Zr-Nb, Ti and Ni. For most materials, the creep coefficient at 340 K was equal to or greater than that at 570 K. Such weak temperature dependence is not easily reconciled with existing models of irradiation creep based on dislocation climb, such as SIPA or climb-induced glide. Rate theory calculations indicate that because the vacancy mobility becomes very low at the lower temperature, recombination should dominate point defect annealing, resulting in a very low creep rate compared to that at the higher temperature. It is shown that the weak temperature dependence observed experimentally cannot be accounted for by the inclusion of more mobile divacancies in the calculation.     

Incompatibility between zircaloy-2 and inconel X-750 during temperature transients

The incompatibility of Zircaloy-2 and Inconel X-750 has been investigated between 1000°C and 1200°C (1200°C being the currently allowable maximum temperature in the acceptance criteria for ECCS for water reactors). It has been found for the temperatures of 1000°C and 1200°C that oxide thicknesses of 2 and 30 μ respectively protect the Zircaloy-2 against attack by Inconel X-750.     

Void formation in solution-treated aisi 316 and 321 stainless steels under 46.5 mev ni6+ irradiation

The swelling and radiation damage structure developed in solution-treated 316 and 321 stainless steels bombarded by 46.5 MeV Ni⁶⁺ ions in the Variable Energy Cyclotron (VEC) have been determined. Foils were pre-injected with 10^?5 a/a He at room temperature and subsequently bombarded by Ni⁶⁺ ions in the temperature range 450–750°C at a damage rate of 1–3 × 10^?3 dpa per second to doses up to 300 dpa and specimens from the foils were examined by transmission electron microscopy. The data obtained were compared with data from other experiments aimed at simulating the fast-neutron irradiation of 316 and 321 steels, in particular previous work with 20 MeV C²⁺ ions and with data on fast-reactor bombarded material. The swelling rates in Ni-ion bombarded specimens were about a factor two less than those in C-ion bombarded specimens and in good agreement with swelling rates in 5 MeV Ni⁺- and neutron-bombarded material. The peak swelling temperature after a dose of 40 dpa was 650°C in 316 steel and 625°C in 321 steel where the swelling was about 5.8% and 4.6% respectively.     

Irradiation creep relaxation of void swelling-driven stresses

Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 °C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 °C and 504 °C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are analyzed using a recently proposed multiaxial creep-swelling model.     

A SANS investigation of the irradiation-enhanced α–α′ phases separation in 7–12 Cr martensitic steels

Five reduced activation (RA) and four conventional martensitic steels, with chromium contents ranging from 7 to 12 wt%, were investigated by small angle neutron scattering (SANS) under magnetic field after neutron irradiation (0.7–2.9 dpa between 250 and 400 °C). It was shown that when the Cr content of the b.c.c. ferritic matrix is larger than a critical threshold value (∼7.2 at.% at 325 °C), the ferrite separates under neutron irradiation into two isomorphous phases, Fe-rich (α) and Cr-rich (α^′). The kinetics of phase separation are much faster than under thermal aging. The quantity of precipitated α^′ phase increases with the Cr content, the irradiation dose, and as the irradiation temperature is reduced. The influence of Ta and W added to the RA steels seems negligible. Cold-work pre-treatment increases slightly the coarsening of irradiation-induced precipitates in the 9Cr–1Mo (EM10) steel. In the case of the low Cr content F82H steel irradiated 2.9 dpa at 325 °C, where α^′ phase does not form, a small irradiation-induced SANS intensity is detected, which is probably due to point defect clusters. The α^′ precipitates contribute significantly to the irradiation-induced hardening of 9–12 wt% Cr content steels.     

Zircaloy-2 irradiation creep strain measurement techniques

The extension measuring devices used in the various in-reactor rigs associated with the UK Zircaloy-2 creep programme are described. All tests were performed at temperatures of about 300° C and fast flux levels (> 1 MeV) in the range $\text{(3–6) × 10}{}^{17}\text{n/m}{}^2\text{/sec}$. The creep rates and total creep strains of uniaxial specimens have been measured using gas gauging techniques. Low creep rates in the range 10^?7 to 10^?5/h have been determined using a conventional type of gauge to an accuracy of within 10% provided that the specimen is on test for a sufficiently long time. For the measurement of high strains (> 1%) a tapered needle gas gauge has been used; reliable strain measurements have been obtained using this device at ambient temperature but at elevated temperature the readings were sometimes found to be in error. The diametral creep strain of small tubes has been determined to within 0.5% using a neutron radiograph technique at reactor shutdown. Provided total strains in excess of a few per cent are expected creep curves can be reliably determined using this method of strain measurement.     

Void Swelling of Solution-Annealed Type 316 Stainless Steel under Long Time Proton Irradiation

Solution-annealed type 316 stainless steel was irradiated by 150 keV proton to a dose of about 6 dpa at the irradiation temperature ranging 450–700°C. To examine the effect of aging during irradiation, the present proton irradiation was carried out for about 25 h at a low dose rate of 7×10–^?5dpa/s. The specimens without He preinjection showed much smaller void swelling than those preinjected with He to the content of 10 at.ppm. Similarly to the case of neutron irradiations, the void swelling in the He preinjected specimens showed the temperature dependence with double peaks, and the peak swelling temperatures were about 550 and 650°C. In these specimens with He preinjection. void number density decreased and average void diameter increased with the increase of irradiation temperature in the range of 450–600°C, but these trends were reversed between 600 and 650°C. The volume of the grain boudary M₂₃C6 precipitates increased with the increase of irradiation temperature from 600 to 700°C, and it was concluded that the decrease of soluble carbon due to the precipitation of M₂₃C6 caused the second swelling peak at 650°C.     

The temperature dependence of nickel-ion damage in nickel

The temperature dependence of void and dislocation structures was studied in high-purity nickel irradiated with 2.8 MeV ⁵⁸Ni⁺ ions to a displacement density of 13 displacements per atom (dpa) at a displacement rate of $\text{7 × 10}{}^{\mathrm{?2}}\text{dpa/sec}$ over the temperature range 325 to 625°C. Dislocation loops, with no significant concentrations of voids, were observed in specimens irradiated at 475°C and below. Specimens irradiated between 525 and 725°C contained both voids and dislocations. The maximum swelling was measured as 1.2% at 625°C. Analysis of the data by theoretical models for void nucleation and growth indicated that the swelling in the present experiment was principally limited by void growth at low temperatures and by void nucleation at high temperatures. The data were also compared with previously reported neutron and nickel-ion irradiation results.     

Annealing of neutron damage in graphite irradiated and stored at room temperature

Nimonic PE16, a gamma-prime Ni₃(Al,Ti) precipitate-strengthened alloy under consideration for fast reactor structural applications, has been neutron irradiated in three heat treatment conditions: solution treated, aged, and overaged. After irradiation at 600° C to 5.4 × 10²²n/cm² (E > 0.1 MeV), or 27 dpa, specimens were characterized for gamma-prime precipitate stability by transmission electron microscopy. The precipitate microstructures after irradiation reflected the influence of the preirradiation heat treatment; and indeed the precipitate particles present prior to irradiation remained stable. However, additional precipitation occurred during irradiation in each of the specimens examined. The in-reactor gamma-prime precipitation process decorated such microstructural features as voids, dislocations and carbide precipitates. Examples were found in the solution-treated condition where gamma prime in the form of Archimedes' screws had precipitated on climbing screw dislocations. The precipitation behavior observed is compared with predictions from existing models. It is concluded that models for solute diffusion to point-defect sinks and for Ostwald coarsening can account for the observations, but that the models for precipitate stability controlled by cascade dissolution during neutron irradiation do not.     

Radiation-induced stress relaxation in high temperature water of type 316L stainless steel evaluated by neutron diffraction

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 10²⁵ n/m² (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σ_x residual stress (transverse to the weld bead) and σ_y residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10⁻⁶/MPa/dpa). This value was smaller than that of nickel alloy.     

Irradiation-Induced Deformation in Cold-Worked SUS 304 Stainless Steel

SUS 304 stainless steel has been used in the light-water reactors constructed in earlier days, in which irradiation-assisted stress corrosion cracking has drawn increasing attention and tensile residual stress is believed to be one of the major causes. It is, therefore, essential to assess its stress relaxation behavior under irradiation, which can be evaluated from the irradiation creep data, and the effect of cold work on it. Creep experiments under 17 MeV proton irradiation (2x10^?7 dpa/s) at 288°C were conducted for SUS 304 with 5% and 25% cold work (CW). Irradiation creep rate of 5%CW was only slightly larger than that of 25%CW. Stress dependence was almost quadratic in both specimens, in contrast with the linear dependence in cold-worked SUS 316L reported earlier. Stress relaxation under irradiation was found to reflect this quadratic dependence. Martensite is induced by cold-working in SUS 304, not in SUS 316L, and marked difference in its amount was found between 5%CW and 25%CW, despite the small difference in irradiation creep behavior. Thus, the observed quadratic dependence appears to result not directly from the induced martensite itself but from a climb-enabled glide of the tangled dislocations densely formed in the vicinity of martensite phase boundaries.