Test methodologies for determining high temperature material properties of thin walled tubes

This report presents briefly the test methods used, within the in the EERA JPNM Pilot Project TASTE, for defining the tensile and creep material properties relevant to the integrity of nuclear fuel claddings. These properties are challenging to extract from thin walled tubes since the standard test methods use test specimen that require minimum material thicknesses in the order of 10 mm or more. In consequence the thin walled material properties are acquired through a number of testing techniques and evaluation methodologies suitable for the thin walled product form. In this report the different test methods and their data assessment requirements are briefly described. The test methods evaluated here comprises of sub-size (curved specimen) tensile testing (ST) of the cladding tube, micro specimen (dog-bone) tensile testing (MT), Small Punch testing (SP), Segmented Expanding Cone Mandrel tests (SCM), the ring tension (RT) and ring compression (RC) tests and internal pressure testing (IP).JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

Determination of high temperature material properties of 15-15Ti steel by small specimen techniques

This report presents the final results from the round robin test program on thin-walled cladding tubes in the EERA-JPNM pilot project TASTE. The test methods and assessment procedures used for the assessment of 15-15Ti steel are presented in a previous TASTE report 1. In this report the test results from different test types are assessed, compared and evaluated. The collation of results, mainly on tensile properties shows good agreement between tests methods. An open question remains if there is some anisotropy between the axial and the hoop direction of the tubes. Results from ring tension indicate lower strength values than the test performed in the axial direction. However, the ring tension calculated estimates do not take bending and friction into account. Tensile strength estimates from miniature Small punch tests samples (3 mm in diameter and 0.25 mm thick) indicate no anisotropy whereas tests on the full wall thickness (0.45 mm) with larger puncher balls indicate a reduction towards the INR measured tensile strength (Ring Tension) in the hoop direction. The ring compression test estimates based on calibration at room temperature by ENEA showed surprisingly good performance in estimating the tensile strength at higher temperatures despite the complex stress distribution for this type of test. The few tests performed for determining creep properties, i.e. small punch creep tests, were not successful in describing the expected creep properties. The SPC specimen (as was the case for some SP "tensile test") showed premature cracking at a very early stage of the test for the cold worked material. As a whole it seems that the different types of tests complement each other and together gives an overall picture of the strength and ductility of the tube material. The classical tests such as the ring compression test and the ring tension test gave good estimates on the hoop strength whereas the small punch tests seemingly give an estimate for the weaker direction.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

Round Robin into Best Practices for the Determination of Indentation Size Effects

peer reviewe

Basic microstructural characterization of second phases in homogeneous weld joint made of X6CrNiNbN25-20 steel after long-term exposure time at 973 K

New blocks of fossil fuel power plants designed for steam temperatures above 873 K require advanced stainless steels as material for superheater or reheater systems. Weld joints are critical parts in fossil power units. Great attention is paid to the exploitation of new steel grades with higher material properties. In the austenitic steels family, the superior grade is undoubtedly HR3C steel (X6CrNiNbN25-20). A detailed knowledge on stability and microstructure composition during thermal exposure of the weld joints made from HR3C is necessary in order to use them in fossil fuel power plants with ultrasupercritical (USC) and new advanced ultrasupercritical (A-USC) steam parameters. The aim of the paper is to identify critical minor phases in HR3C steel, which allow acceleration of creep damage. The sigma-phase and rough carbides M23C6 type is considered as such a phases in this steel. In this study, the sigma-phase is identified and studied in more detail in relation to the development of creep damage at 973 K. Experimental material of the homogeneous HR3C weld joints in two states: in the as-welded state (AW) and after the postweld heat treatment (PWHT). Weld joints were manufactured by orbital welding using the gas tungsten arc welding (GTAW) method, heat input Q(s) = 1600 J/mm, interpass 423 K, three beads. Nickel-base alloy UTP A6170 Co (equivalent to Thermanit 617) was used as a filler material. The PWHT was carried out at the temperature of 1503 K for 15 min. Stress rupture tests were performed on the cross-weld (CW) joints of tubes o 38 x 6.3 mm at 973 K with times to rupture up to nearly 22,000 h. The polished surface of the longitudinal sections was subjected to color etching in Murakami (30 g K-3(Fe(CN)(6)), 30 g KOH, 60 ml H2O) in order to highlight the sigma-phase. Several microscopic techniques were used for the study. The results were supplemented by creep, grain size, and microhardness data hardness vickers (HV) 0.5. The PWHT specimens exhibited an average sigma-phase size of approximately 5 mu m as well as AW specimens in specimens with short time to rupture (t(r)). However, t(r) such as 20,000 h, the average sigma-phase size already reached dangerous border 10 mu m. The AW specimens as opposed to the PWHT specimens did not show a noticeable growth of austenitic grains in the heat-affected zone (HAZ). In specimens after PWHT, the average grain size in HAZ was more than twice that of the body material (BM). It is worth noting that creep ductility values of specimens in the state after PWHT are very low, which is the result of coarse-grained structure in the HAZ and accelerated precipitation of sigma-phase particles along grain boundaries during creep at 973 K.Web of Science72art. no. 02450

Status of the EU DEMO breeding blanket manufacturing R&D activities

The realization of a DEMOnstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle by 2050, is viewed by Europe as the remaining crucial step towards the exploitation of fusion power. The EUROfusion Consortium, in the frame of the European Horizon 2020 Program, has been assessing four different breeding blanket concepts in view of selecting the reference one for DEMO. This paper describes technologies and manufacturing scenarios developed and envisaged for the four blanket concepts, including nuclear “conventional” assembly processes as GTAW, electron beam and laser welding, Hot Isostatic Pressing (HIP), and also more advanced (from the nuclear standpoint) technologies as additive manufacturing techniques. These developments are performed in conformity with international standards and/or design/manufacturing codes. Topics as the metallurgical weldability of EUROFER steel and the associated risks or the development of appropriate filler wire are discussed. The development of protective W-coating layers on First Wall, with Functionally Graded (FG) interlayer as compliance layer between W and EUROFER substrate, realized by Vacuum Plasma Spraying method, is also propounded. First layer systems showed promising layer adhesion, thermal fatigue and thermal shock properties. He-cooled mock-ups, representative of the First Wall with FG W/EUROFER coating are developed for test campaigns in the HELOKA facility under relevant heat fluxes.First elements of Double Walled Tubes (DWT) manufacturing and tube/plate assembly for the water cooled concept are given, comprising test campaign aiming at assessing their behaviour under corrosion.In addition, further development strategies are suggested

Status of the EU DEMO breeding blanket manufacturing R&D activities

The realization of a DEMOnstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle by 2050, is viewed by Europe as the remaining crucial step towards the exploitation of fusion power. The EUROfusion Consortium, in the frame of the European Horizon 2020 Program, has been assessing four different breeding blanket concepts in view of selecting the reference one for DEMO. This paper describes technologies and manufacturing scenarios developed and envisaged for the four blanket concepts, including nuclear “conventional” assembly processes as GTAW, electron beam and laser welding, Hot Isostatic Pressing (HIP), and also more advanced (from the nuclear standpoint) technologies as additive manufacturing techniques. These developments are performed in conformity with international standards and/or design/manufacturing codes. Topics as the metallurgical weldability of EUROFER steel and the associated risks or the development of appropriate filler wire are discussed. The development of protective W-coating layers on First Wall, with Functionally Graded (FG) interlayer as compliance layer between W and EUROFER substrate, realized by Vacuum Plasma Spraying method, is also propounded. First layer systems showed promising layer adhesion, thermal fatigue and thermal shock properties. He-cooled mock-ups, representative of the First Wall with FG W/EUROFER coating are developed for test campaigns in the HELOKA facility under relevant heat fluxes. First elements of Double Walled Tubes (DWT) manufacturing and tube/plate assembly for the water cooled concept are given, comprising test campaign aiming at assessing their behaviour under corrosion. In addition, further development strategies are suggested