Extension and refurbishment of the CEA Saclay Leci hot laboratory initiated 10 years ago is now considered as completed. The present paper describes two new cladding testing facilities available at Leci. Iodine induced stress corrosion cracking of zirconium alloys fuel claddings may develop during power transient due to pellet-cladding interaction (PCI). This phenomenon was studied in the previous CEA installation using iodine atmosphere internal pressure creep tests on defuelled samples. The new facility allows tests to be performed in two modes: i) driving on the pressure with the ability to do complex loading creep tests, ii) driving on the hoop strain thanks to the diameter laser measurement, which allows relaxation tests to be achieved and consequently to be more representative of the loading imposed by the pellet on the cladding during PCI. Inert atmosphere can also be used in order to determine mechanical properties of irradiated claddings or hydrides reorientation conditions. Axial creep: Two electro-mechanical tensile machines have been set up to test several geometries of irradiated materials: tubular and flat specimens or bars. Variable gauge length extensometers with ceramic rods have been adapted to handling with telemanipulators and are used for both creep and relaxation tests up to 800°C.

The LCMl (Laboratorv for characterization of irradiated materials), located in CEA from Saclav. is in charge oi the mechanical tests on iriaiated materials. The dynamic tensiie testing machine, in a hot celi equipped with two remote handlings. has been first improved in 1995. to fulfiil the French safety programs on Reactivity Initiated Accident C A 0 1 .::U SCI .5 <l :11 S ~ 2 . n rr i to oni.3 11 rwt;ua,~cs cropcn, cnla on r..rrrrl L ,:a", c m " ng $,p4 ' l r U e u l i q . m t t , in,: -;u~ o j s ieapmsc .noer H A or .OCA lrios ert cscnq 3n.1 lnc!rnn coroton; Fc' lnt HIA. 1.1 2 illem, 1 2 i l ?g 1 wdr8 r.,'e, .p 1, 5 i' a m nsal ng rd!es .p !o 210.C S . ;n F 1% _OS? 01 CS dnl ACC odi i s . OCA led g ;: sra r i.?lr, U. 10 an0 n c x r g rales u. ?WC s KO. o oi. a00102r .?h: The tensile samples are machined with a spark erosion machine. directiy from pieces of cladding previously defueied. Two kinds of samples can be machined in the ciadding: Axial samples in order to test axial mechanical characteristics Rina samoles in order to test transverse mechanical characteristics, more reoresentative of HIA On one hand. the axial tensile tests were performed using the Jouie effect, and heating rates up to about 500°C.S' were obtained. This enabled us to perform the axiai tests in a satisfactory manner. On the other hand, the tensiie ring tests were first performed in a vertical furnace with a heating rate about 09T .s ' and a thermal stability about 1°C. For temperatures above 480°C. the mechanical characteristics showed a sharp drop which could be attributed to irradiation defect annealing. Therefore we have recently deveiaped an Induction heating system to reach heating rates high enough (200'c.s~') to prevent any significant anneaiing before performing the ring tensile tests. To apply a uniaxiai tangential tension, two matching half-cylinders are inserted inside the ring and are pulled apait. The main objective of this paper is to present this system that can be telemanipulated and achieve heatlng rates up to ~ O O ~ C . ~ . ' while taking into account the requirement forair-cooled coils in the hat celi. The same interface is 40 Aver@ ei ar /lnduclion heating on dynamic lensile rests in CEA Saclay used for mduction heating equipment and Joule effect (current system), in order to control the specimen temperature synchronously with the Iaadlextension, colleCting data from all transducers connected including load/displacement.

The hardening and the embrittlement under neutron irradiation of an A508 type RPV steel considering three different microstructures (bainite, bainite-martensite and martensite)have been investigated These microstructures were obtained by quenching after autenitization at 1100 degrees C. The irradiation induced hardening appears to depend on microstructure and is correlated to the yield stress before irradiation. The irradiation induced embrittlement shows a more complex dependence. Martensite bearing microstructures are more sensitive to non hardening embrittlement than pure bainite. This enhanced sensitivity is associated with the development of intergranular brittle facture after irradiation; the pure martensite being more affected than the bainite-martensite. It is of interest to note that this mixed microstructure appears to be more embrittled than the pure bainitic or martensitic phases in terms of temperature transition shift. This behaviour which could emerge from the synergy of the embrittlement mechanisms of the two phases needs further investigations. However, the role of microstructure on brittle intergranular fracture development appears to be qualitatively similar under neutron irradiation and thermal ageing. (C) 2015 Elsevier B.V. All rights reserved.

Four 9Cr Reduced Activation martensitic steels, two ODS-FeCr alloys and one conventional 9Cr martensitic steel were irradiated in BOR-60 reactor to doses ranging from 40 to 78dpa at 325°C. Tensile, impact tests and fracture surface examinations were subsequently carried out. All 9Cr steels showed strong increases in yield stress, decreases in ductility and large shifts of the Ductile-to-Brittle transition temperatures, however the degradation of tensile and impact properties was much more pronounced for the conventional 9Cr–1MoVNb steel. This steel displayed a brittle tensile behavior above 40 dpa of exposure whereas EUROFER retained significant ductility up to 78dpa. MA957 ODS in the fine-grain non-recrystallised condition, exhibited moderate irradiation-induced hardening together with a ductile behavior up to the highest dose, but its impact properties after irradiation were inferior to those of 9Cr RA steels, while EUROFER ODS irradiated to 40dpa showed poor tensile and impact behavior.

In the European EUROTRANS/DEMETRA program, the synergistic effect of radiation damage and helium on microstructure and mechanical properties of two 9Cr 1Mo ferritic/martensitic (FM) steels T91 and EM10 was evaluated after irradiation in SINQ targets. In addition, the helium induced effect was investigated using helium implanted specimens. The results demonstrate that helium can induce significant embrittlement effect in FM steels as shown by the tremendous increase in ductile-to-brittle transition temperature, the great reduction in ductility and fracture toughness at>∼15dpa and 1000appm He and the occurrence of intergranular fracture mode. Further, high-density helium bubbles can produce pronounced hardening effect.

Reconstitution techniques are often used to allow material from previously fractured Charpy-V specimens to be reused for additional experiments. This paper presents a comparative experimental study of various reconstitution techniques and evaluates the feasibility of these methods for future use in shielded cells. The following techniques were investigated: arc stud welding, 6.0 kW CO(2) continuous wave laser welding, 4.5 kW VAG continuous wave laser welding and friction welding. Subsize Charpy specimens were reconstituted using a 400W YAG pulsed wave laser. The best result was obtained with arc stud welding; the resilience of the reconstituted specimens and the load-displacement curves agreed well with the reference specimens, and the temperature elevation caused by the welding process was limited to the vicinity of the weld. Good results were also obtained with friction welding: this process led to the best quality welds. Laser welding seems to have affected the central part of the specimens, thus leading to different resilience values and load-displacement curves. (C) 2011 Elsevier B.V. All rights reserved.

Solution annealed 304L (SA 304L) and cold work 316 (CW 316) austenitic stainless steel irradiation creep behaviour have been studied thoroughly. Irradiations were carried out in fast breeder reactors BOR-60 (at 330 degrees C, up to 120 dpa) and EBR-II (at 375 degrees C, up to 10.5 dpa), and in the OSIRIS mixed spectrum reactor (at 330 degrees C, up to 9.8 dpa). After an incubation threshold, the irradiation creep of the austenitic stainless steels is linear in stress and in dose. Creep appears to be athermal in this temperature range. A significant difference in the behaviour is measured between the creep of SA 304L and CW 316.In order to study the anisotropy of loop population, which would be the signature of a possible stress induced preferential absorption (SIPA) mechanism for irradiation creep, special attention was given to the measurement of anisotropy of loop distribution between the four families. The anisotropy induced by an applied stress has been shown to be in the range of the statistical scatter in the situation where no stress is applied. TEM microstructural analyses performed on this sample show slight difference between the microstructure of specimens deformed under irradiation and the microstructure of specimens irradiated without stress under the same irradiation conditions. (C) 2011 Elsevier B.V. All rights reserved.

In blankets of a fusion power plant 14MeV neutrons produce displacement damage, dpa, and helium and hydrogen. Martensitic steels offer the advantages of low swelling and reduced helium embrittlement compared to austenitic steels. Reduced activation Eurofer 97 steel has been exposed to neutron displacement damage up to about 15dpa in materials test reactors such as OSIRIS and HFR at temperatures in the range of RT to 600°C, and up to 80dpa in BOR-60 at temperatures in the range of 300–330°C. The post-irradiation mechanical properties in the range of 300–330°C show increases in yield stress, decrease in ductility and an increasing ductile to brittle transition temperature. The hardening rate is decreasing with increasing damage level, but it does not show saturation at doses examined. Analyzing the present results and reviewing properties of other steels with different compositions, including ODS steels, lead to the conclusion that improvement of the radiation resistance of steel will be based on nano-microstructural features.

The anisotropic plastic behavior and the fracture of as-received and hydrided Cold-Worked Stress Relieved Zircaloy-4 cladding tubes are investigated under thermal–mechanical loading conditions representative of Pellet–Clad Mechanical Interaction during Reactivity Initiated Accidents in Pressurized Water Reactors. In order to study the combined effects of temperature, hydrogen content, loading direction and stress state, Axial Tensile, Hoop Tensile, Expansion Due to Compression and hoop Plane Strain Tensile tests are performed at room temperature, 350°C and 480°C on the material containing various hydrogen contents up to 1200wt.ppm (hydrides are circumferential and homogeneously distributed). These tests are combined with digital image correlation and metallographic and fractographic observations at different scales. The flow stress of the material decreases with increasing temperature. The material is either strengthened or softened by hydrogen depending on temperature and hydrogen content. Plastic anisotropy depends on temperature but not on hydrogen content. The ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking. The plastic strain that leads to hydride fracture at room temperature decreases with increasing hydrogen content. The influence of stress triaxiality on hydride cracking is negligible in the studied range. The influence of hydrogen on material ductility is negligible at 350°C and 480°C since hydrides do not crack at these temperatures. The ductility of the material increases with increasing temperature. The evolution of material ductility is associated with a change in both the macroscopic fracture mode of the specimens and the microscopic failure mechanisms.

Ferritic-martensitic steels are prime candidate materials for future reactors. We present here the results of a study on the effects of helium implantation on the fracture behavior of 9Cr (T91) martensitic steels. Three-points static bending tests were performed at room temperature on implanted specimens and at −170°C on un-implanted material. All these tests led to brittle fracture. Based on a mechanical analysis of the tests results using Finite Element calculations, we have proposed that the mechanism of brittle fracture is controlled by a double criterion depending on implantation temperature and helium content. Furthermore, by applying the Beremin model, the toughness of helium implanted steel has been evaluated.

The anisotropic viscoplastic behavior and the fracture of cold-worked stress relieved Zircaloy-4 cladding tubes is investigated under reactivity initiated accidents loading conditions. The combined effects of temperature (from 25°C up to 480°C), hydrogen content (from 0 up to 1200 ppm) and stress/stain state (from uniaxial tension up to plane strain tension) are analyzed. A strengthening effect of hydride precipitates and a softening effect of dissolved hydrogen are observed. Ductility of the material increases with increasing temperature and decreases with increasing hydrogen content at room temperature. The embrittlement effect of hydrides is substantially reduced when increasing temperature. A slight effect of stress/strain triaxiality is evidenced. The evolution of material ductility is associated with a change in the macroscopic fracture aspect and the failure mechanisms. A model is proposed to describe the anisotropic viscoplastic behavior of the material. Introduction In order to optimize the use of nuclear fuels and improve the management of Pressurized Water Reactors (PWRs), it is planned to progressively extend the discharge burnup of fuel assemblies. This extension would enhance cladding degradation and may increase the likelihood of cladding failure during postulated design basis accidents such as the Reactivity Initiated Accident (RIA). Caused by the inadvertent ejection of a control rod, this scenario leads to an abrupt fuel pellet thermal expansion, which produces a Pellet-Clad Mechanical Interaction (PMCI). This PCMI forces the fuel cladding to deform at high strain rate (typically 1 s) under multiaxial tension (axial and hoop components) with a strain path between plane strain hoop tension and equal-biaxial tension [1]. During this early stage of the transient, temperature of the cladding approximately ranges from 350°C up to 600°C. During reactor operation, the metal-water reaction at the fuel cladding outer surface introduces hydrogen into the metal. Hydrogen uptake is commonly identified as a main contributor to limiting the fracture resistance of high burnup fuel claddings during the PCMI stage of RIAs [2]. Once its solubility limit is reached, hydrogen precipitates as -hydride platelets and reduces – depending on temperature – ductility of the cladding material [3-6]. Furthermore, it is expected that the stress state influence the mechanical behavior of the hydrided material [3]. The present work addresses the characterization and the modeling of the combined effects of hydrogen, temperature and stress/strain state on both the viscoplastic behavior and the failure of Zircaloy-4 fuel claddings under RIA loading conditions. Experiment Material. The material of this study is Cold-Worked Stress Relieved Zircaloy-4 tube, which is commonly used for fuel claddings in french PWRs. Its chemical weight composition is 1.3% Sn,

This paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of cold-worked stress relieved (CWSR) Zircaloy-4 fuel claddings submitted to reactivity initiated accident (RIA) loading conditions. The model relies on a multiplicative viscoplastic formulation and reproduces strain hardening, strain rate sensitivity and plastic anisotropy of the material. It includes temperature, fluence and irradiation conditions dependences within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially obtained from the PROMETRA program, dedicated to the study of zirconium alloys under RIA loading conditions. Once calibrated, the model provides a reliable description of the mechanical behavior of the fresh and irradiated (fluence up to 10×1025nm-2 or burnup up to 64 GWd/tU) material within large temperature (from 20°C up to 1100°C) and strain rate ranges (from 3×10-4s-1 up to 5s-1), representative of the RIA spectrum. Finally, the model is used for the finite element analysis of the hoop tensile tests performed within the PROMETRA program.

Tensile specimens of 9Cr–1Mo (EM10) and mod 9Cr–1Mo (T91) martensitic steels in the normalized and tempered metallurgical conditions were irradiated with high energy protons and neutrons up to 20 dpa at average temperatures up to about 360 °C. Tensile tests were carried out at room temperature and 250 °C and a few samples were tested at 350 °C. The fracture surfaces of selected specimens were characterized by Scanning Electron Microscopy (SEM). While all irradiated specimens displayed at room temperature considerable hardening and loss of ductility, those irradiated to doses above approximately 16 dpa exhibited a fully brittle behaviour and the SEM observations revealed significant amounts of intergranular fracture. Helium accumulation, up to about 0.18 at.% in the specimens irradiated to 20 dpa, is believed to be one of the main factors which triggered the brittle behaviour and intergranular fracture mode. One EM10 and one T91 specimen irradiated to 20 dpa were annealed at 700 °C for 1 h following irradiation and subsequently tensile tested. In both cases, a remarkable recovery of ductility and strain-hardening capacity was observed after annealing, while the strength remained significantly above that of the unirradiated material.

The tensile properties of mechanically-alloyed oxide dispersion strengthened MA957 steel were measured at room temperature following irradiation in the SINQ spallation target up to almost 20dpa corresponding to an accumulated helium content of about 1750appm, with an average irradiation temperature range of 100–360°C. In contrast to the behaviour of 9Cr–1Mo martensitic steel samples subjected to identical irradiation conditions and which were drastically embrittled at high dose, all tested MA957 specimens displayed a ductile fracture mode as shown by the measured values of uniform and total elongations and by the results of fracture surface examinations. This good mechanical behaviour is a new evidence that this type of material may be able to sustain high displacement damage and helium levels and is thus particularly well suited for fusion applications.

An assessment of the mechanical properties of the highly irradiated fuel claddings under high strain rate has been carried out in the framework of the PROMETRA program undertaken by the French Institut de Radioprotection et de Surete Nucleaire in collaboration with Electricite de France and Commissariat a l' Energie Atomique (CEA). Three types of tests, including burst tests, hoop and axial tensile tests, have been performed at CEA Saclay hot laboratories to determine the cladding tensile properties to use in the SCANAIR code. The prototypicality of each test with regard to the reactivity-initiated accident loading conditions can be addressed and analyzed in terms of strain or stress ratio. The high-strain-rate ductile mechanical properties of irradiated ZIRLO and M5 alloys derived from the PROMETRA program and their comparison to the stress-relieved irradiated Zircaloy-4 are reported. Then, the clad brittle behavior, in particular for highly corroded or spalled Zircaloy-4 cladding, is investigated.

An experimental characterization was conducted of helium effects on the mechanical properties of a 9Cr martensitic steel. Six sub-size Charpy samples were implanted in the notch region at 250°C with 0.25at.% helium and subsequently tested in 3-point bending at room temperature. Brittle fracture mode (cleavage and intergranular fracture) was systematically observed in the implanted zones of the samples. Finite element calculations of the tests, using as input the tensile properties measured on a helium loaded sample, were performed in order to determine the fracture stress at the onset of brittle crack propagation. Preliminary TEM investigations of the implantation-induced microstructure revealed a high density of small helium bubbles.

An experimental irradiation, named “Alexandre,” has been carried out in the Osiris experimental reactor to perform a generic study on the mechanical behavior after irradiation at 325°C of different kinds of steels suitable for use as irradiated components in a nuclear reactor [1]. The irradiated steels were austenitic stainless, martensitic (conventional and reduced activation), and ferritic-martensitic Oxide Dispersion Strengthened steels in various initial metallurgical conditions. The final dose was 9 dpa, which represents nearly a “saturation” dose for the hardening/embrittlement of both austenitic and martensitic steels. At this dose, the Yield Strength and the Ultimate Tensile Strengths are almost equal, and strong localization of the plastic deformation is often observed. After irradiation, as-quenched martensitic steels exhibit very large tensile strengths, and some of them show ductility parameters comparable with those of the tempered martensitic steels. At 9 dpa, the behavior of cold-worked steels tends to be similar to that of tempered steels with a more pronounced localization of the deformation. This indicates that the recovery of the cold-worked microstructure is not achieved at 9 dpa. Certain martensitic steels (MANET II and HT9) show tendencies to brittle behavior and exhibit a considerable degradation of their ductility parameters after irradiation. The behavior of the most chromium-rich ferritic steel (ODS-MA957) is quite surprising. It exhibits moderate hardening and good post-irradiation ductility parameters. Measurements of the dose rate have been performed after several cooling-times, of up to 53 months after unloading, to study the radioactive decay of the steels. Experimental data show that the reduced activation steels exhibit the lowest residual activity and the highest rate of relative activity decrease (90 %) during the cooling period.

Martensitic/ferritic steels (containing 7–13 % Cr) are candidate materials for internal structures in pressurized water, fast breeder, and fusion reactors. Approval for use requires verification of structural stability under neutron irradiation in relation to the evolution of mechanical properties. In this context, several conventional and Reduced Activation (RA) martensitic materials were neutron irradiated at 325°C up to 6 dpa. They were investigated by Small Angle Neutron Scattering (SANS) under a magnetic field after various doses. It was shown that when the Cr content of the b.c.c. ferritic matrix was larger than a critical threshold value (∼ 7.2 at.% at 325°C), the ferrite separated under neutron irradiation into two isomorphous phases, Fe-rich (α) and Cr-rich (α′). The kinetics of phase separation is much faster than under thermal aging. The quantity of precipitated α′ phase increases with the Cr content and the irradiation dose. In the case of steel with the lowest Cr content (F82H) irradiated at 5.6 dpa at 325°C, the α′ phase does not form, and the SANS signal suggests a small contribution due to vacancy clusters. It was believed that these could contribute to the “black dots” observed by TEM. Furthermore, we studied the microstructural features responsible for the secondary hardening phenomenon detected in the as-quenched F82H martensitic steel during irradiation or annealing. In addition, the microstructural evolution of the Oxide Dispersion Strengthened (ODS) steel MA957, which presents an excellent hardening/ductility compromise after irradiation, has been also characterized. The stability of the oxides has been elucidated, and an important α′ volume fraction has been detected. The contribution of α′ to the irradiation-induced hardening was assessed. This, although not negligible, is not the critical factor in normalized and tempered or cold-worked steels. However, it may be the main contribution to hardening in MA957.

In order to investigate helium effects on the fracture properties of martensitic mod 9Cr–1Mo (T91) steel, miniature Charpy specimens were implanted at 250°C in the notch region to 0.25at.% helium using a degraded 34MeV 3He ion beam and subsequently submitted to static bending tests at room temperature. For the six implanted specimens, a ‘pop-in’ phenomenon, which is an arrested unstable crack extension, was systematically recorded during testing. In the implanted zones of the samples, the fracture mode was fully brittle with both intergranular and cleavage fracture, whereas for unimplanted samples tested at −170°C, the fracture mode was found to be 100% cleavage. Finite element simulations of the tests performed on unimplanted and implanted specimens were also carried out to determine stress and strain fields at the onset of crack propagation. Based on these computations, the fracture toughness of implanted T91 was tentatively evaluated using the Beremin model of the local approach to brittle fracture.