[en] After 17 years of operation, a leak was detected in one of the two spare Control Element Drive Mechanism (CEDM) Upper Housings on the reactor vessel closure head of a pressurized water reactor. The housings contained primary water coolant under high temperature and pressure. Immediately after discovery of the leak, the reactor had to be placed in a cold shutdown condition to allow investigation into the cause of the cracking. The forced plant outage lasted four weeks and cost several million dollars in lost revenue. Metallographic examination of the failed spare housing found that the through-wall crack was oriented axially. A second smaller axial crack was also found in the failed housing. Both cracks were localized to the vicinity of a weld overlay are of the SA-312 type 348 austenitic stainless steel housing. Based on the similarity of operations in both spare housings, the other intact spare housing was also removed for examination, and similar axial cracking was observed on the inside diameter

No abstract available

[en] Annealed type-348 stainless steel specimens irradiated to 33 to 39 dpa at 350⁰C were examined by transmission electron microscopy to determine the cause of pronounced irradiation creep and hardening. The irradiation produced very high densities of 1-2 nm diameter helium bubles, 2-20 nm diameter faulted (Frank) dislocation loops and 10 nm diameter precipitate particles. These defects account for the observed irradiation hardening but do not explain the creep strains. Too few point defects survive as faulted dislocation loops for significant creep by the stress-induced preferential absorption (SIPA) mechanism and there are not enough unfaulted dislocations for creep by climb-induced glide. Also, the irradiation-induced precipitates are face-centred cubic G-phase (a niobium nickel silicide), and cannot cause creep. It is suggested that the irradiation creep occurs by a grain-boundary movement mechanism such as diffusion accomodated grain-boundary sliding. (orig.)

[en] In-reactor creep and swelling of Type 248 stainless steel from ATR SN-5 and ETR H-10 in-pile tube measurements were investigated to identify and characterize their mechanistic relationships at temperatures less than 723⁰K. The principal creep mechanism appears to be diffusion along high conductivity paths related to interstitial loops. The irradiation creep is a function of temperature and is presented as an empirical equation. The swelling in the ATR in-pile tubes is also presented as an empirical equation

[en] This report presents the results of a post-irradiation destructive examination of a Type 348 stainless steel inpile pressure tube, which was removed from the Advanced Test Reactor after exposure to a peak fluence of 3.2 x 10²² n/cm² (E > 1.0 MeV). This in-core portion of the tube operated under a hoop stress of 10,500 psi at a mean wall temperature of 725⁰F. The test results indicate the occurrance of radiation saturation of the tensile, elongation, and fracture toughness properties, a reduced but adequate 800⁰F creep-rupture strength, and the absence of 800⁰F notch sensitivity. Radiation induced changes in density and thermal diffusivity were minimal. The material incurred an in-reactor creep of 1.8% and a concomitant stress relaxation. Transmission electron microscopy also indicates that radiation-induced lattice damage had achieved saturation

[en] Concerns for irradiation-assisted stress corrosion cracking (IASCC) of reactor internals are increasing, especially for components that are not readily replaceable. Both laboratory and field data show that intergranular stress corrosion cracking of stainless steels and nickel-base alloys can result from long term exposure to the high energy neutron and gamma radiation that exists in the core of light water reactors (LWR's). Radiation affects cracking susceptibility via changes in material micro-chemistry (radiation induced segregation, or RIS), water chemistry (radiolysis) and material properties/stress (e.g., radiation induced creep and hardening). Based on many common dependencies, e.g., to solution purity, corrosion potential, crevicing and stress, IASCC falls within the continuum of environmental cracking phenomenon in high temperature water

No abstract available

[en] After the Fukushima nuclear accident in 2011, the nuclear research community has initiated research into the development of fuels that are resistant to accidents. In this context, iron-based alloys have emerged as a good alternative to zirconium alloys. In order to make possible the cladding material replacement, studies related to their mechanical properties are necessary. Thus, the present study carried out a mechanical-structural evaluation from the available data collection regarding the mechanical properties of stainless steel 348, specifically in the conditions of the burst test. Burst tests were performed at various temperatures ranging from 32°C up to 450 °C. Then, a computational model was created based on the specimen of the burst test. Numerical simulation was performed considering the tensile tests of stainless steel at various temperatures. The numerical results were compared with the results of the burst test. Test and simulations were comparable leading to computational model validation. As austenitic stainless steels have structural stability for low and high temperatures, the results could be extrapolated to temperatures higher than those in the burst test. After the validation of the computational model, simulations were performed for temperatures higher than 450ºC, thus obtaining a burst pressure curve as a function of the temperature for stainless steel ANSI 348. The correlation of burst data as function of temperature could be implemented in the FRAPTRAN code, in order to make possible the evaluation of the behavior of a fuel rod with stainless steel ANSI 348 under postulated accident conditions (LOCA). (author)

[en] The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

[en] A wide range of industries including energy, chemistry, pharmacy, textiles, food and drink, pulp and paper, etc. is using stainless steels. Metastable austenitic steels such as used in power plants and chemical industry are subjected to cyclic mechanical and thermal loading in air as well as under the influence of corrosive media. This paper provides an overview on different nondestructive and electrochemical measurement techniques, which allow differentiating fatigue damage effects in total strain controlled multiple and constant amplitude tests with respect to damage appearance on surface, in subsurface area as well as in volume of specimens or components microstructure. In addition to conventional mechanical stress-strain hysteresis curves, electrical resistance, magnetic and open circuit potential measurements have been applied to characterize the cyclic deformation behavior of the metastable austenitic steel AISI 348 (X10CrNiNb18-9) in laboratory air and in distilled water. Based on these results obtained, the paper provides an outlook on the possibility for an efficient (remaining) fatigue life evaluation approach, which is adapted to the needs of the application areas.