Creep rupture strength of heat affected zone for 9Cr ferritic heat resisting steels

This paper describes creep rupture characteristics of weld heat affected zone, HAZ for 9Cr ferritic steels that are promising materials for nuclear energy uses. In general, creep rupture strength in the heat affected zone of peak temperature between 900 and 1000°C is lower than that in the base metal for ferritic steels. Grain refinement and coagulation of carbides for 9Cr–1Mo steels cause decrease in creep rupture strength of the HAZ. The hardness in the simulated HAZ heated to around 1000°C decreases during creep. This seems to be related to weakening of the HAZ at 1000°C. However, substitution of W for Mo is very effective in enhancement of creep rupture strength of the HAZ due to higher stability of carbides and increase in quantity of precipitated carbides during creep rupture test.     

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.     

Mechanical property characterization in support of elevated temperature reactor design

The mechanical properties data used for the design of nuclear reactors and reactor components for operation at elevated temperatures often show considerable variation. In types 304 and 316 stainless steels, much of the heat-to-heat variation is due to variation in chemical composition and grain size. For a heat-treatable ferritic steel such as 2 1/4 Cr-1 Mo steel, chemical composition is again important, but the heat treatment is more important. An understanding of the metallurgical processes that lead to data variation and property changes in these steels can be used to optimize reactor design.     

Nuclear material investigations by advanced analytical techniques

Advanced analytical techniques have been used to characterize nuclear materials at the Paul Scherrer Institute during the last decade. The analysed materials ranged from reactor pressure vessel (RPV) steels, Zircaloy claddings to fuel samples. The processes studied included copper cluster build up in RPV steels, corrosion, mechanical and irradiation damage behaviour of PWR and BWR cladding materials as well as fuel defect development. The used advanced techniques included muon spin resonance spectroscopy for zirconium alloy defect characterization while fuel element materials were analysed by techniques derived from neutron and X-ray scattering and absorption spectroscopy.     

Monte Carlo calculation of radiation damage in first wall of an experimental hybrid reactor

This work was focused on the neutronic calculation of the nuclear parameters (neutron spectrum, displacement per atom (DPA), gas production, tritium breeding ratio (TBR), nuclear heating) for structural materials in the first wall (FW) and fuel clad (made of ferritic/martensitic steels, vanadium alloy, silicon carbide, copper alloy, and stainless steel) of an experimental hybrid reactor using the most current Monte Carlo Neutron-Particle Transport code MCNP5 1.4. Neutronic calculations were performed using a (DT) fusion driver hybrid reactor under a neutron wall loud of 2.25 MW/m² by full reactor power for one year. Obtained results were compared with three different data libraries (ENDF/B-V, ENDF/B-VI and CLAW-IV). TBR values in the reactor blanket for all investigated materials became greater than the minimum requirement (TBR > 1.05). Nuclear parameters like DPA, He-production and nuclear heating were considered as radiation damage limits for structural materials, copper alloy (Cu0.5Cr0.3Zr) showed better performance than all investigated materials.     

Presentation of the properties of the ferritic alloy 2.25 Cr-1 MoNiNb for steam generator application in FBR systems

2.25 Cr-1 MoNiNb steel, used for the construction of the steam generator of a fast breeder reactor, is subjected to operation at elevated temperature in the creep range. Although this operation condition is a limiting factor for allowable loads during normal operation, it is necessary to have sufficient knowledge of the strength-properties of this kind of steel after thermal aging. Experiences with this steel are described. It is shown that this stabilized ferritic steel reveals the common behaviour of all ferritic steels at elevated temperature. The change of mechanical properties can easily be quantified. Special attention was given to the analysis of strain-controlled tensile testing and the uniform elongation of s.g.-tubing.     

Effects of Grain Morphology and Texture on High Temperature Deformation in Oxide Dispersion Strengthened Ferritic Steels

Oxide Dispersion Strengthened (ODs) ferritic steel is the promising candidate alloy for long-life core materials of fast reactor. A series of experiments, such as tensile tests, creep rupture tests, texture measurements and microstructure observations, are performed for the fabricated sheets of ODs ferritic steel with simulating type of morphology and also for the cladding tube in order to clarify the origin of the peculiar strength anisotropy of the cladding tube: degraded creep rupture strength in hoop direction. From these experiments, effects of grain morphology and texture on deformation of ODs ferritic steels are evaluated.

The sheets and the cladding tube have strong texture of {001}?110? and {111}?110?, respectively. In longitudinal and transverse directions of the sheets, strength level is significantly different from each other, but crystallographic orientation is almost equivalent. From that finding, it is considered that strength anisotropy of the cladding tubes is not attributed to the texture. From the results of micro structure analysis, it is concluded that origin of the degraded creep rupture strength in transverse hoop direction of the cladding tube comes from the grain boundary sliding at the large tilt angles.     

Fe-13 Cr-Ti-Mo ferritic alloy as material for a fast reactor

Recent experimental work tends to indicate that embrittlement and swelling due to radiation by fast neutrons are not so important in ferritic as in austenitic stainless steels. For this reason, a Fe-13 Cr-Ti-Mo ferritic alloy has been developed as material for a fast reactor. This alloy is compatible with a sodium environment and derives its mechanical properties from solid-solution hardening and χ-phase precipitation. Mechanical properties have been investigated on this ferritic alloy in the cold rolled and annealed conditions. Stress rupture properties at 550°C have been compared with those of two commercial ferritic steels namely the X 20 Cr Mo (W) V and X 10 Cr 13 steels.     

Corrosion-resistant coating technique for oxide-dispersion-strengthened ferritic/martensitic steel

Oxide-dispersion-strengthened (ODS) steels are attractive materials for application as fuel cladding in fast reactors and first-wall material of fusion blanket. Recent studies have focused more on high-chromium ferritic (12–18 wt% Cr) ODS steels with attractive corrosion resistance properties. However, they have poor material workability, require complicated heat treatments for recrystallization, and possess anisotropic microstructures and mechanical properties. On the other hand, low-chromium ferritic/martensitic (8–9 wt% Cr) ODS steels have no such limitations; nonetheless, they have poor corrosion resistance properties. In our work, we developed a corrosion-resistant coating technique for a low-chromium ferritic/martensitic ODS steel. The ODS steel was coated with the 304 or 430 stainless steel, which has better corrosion resistances than the low-chromium ferritic/martensitic ODS steels. The 304 or 430 stainless steel was coated by changing the canning material from mild steel to stainless steel in the conventional material processing procedure for ODS steels. Microstructural observations and micro-hardness tests proved that the stainless steels were successfully coated without causing a deterioration in the mechanical property of the low-chromium ferritic/martensitic ODS steel.     

Stability of nanometer-sized oxide clusters in mechanically-alloyed steel under ion-induced displacement cascade damage conditions

Oxide dispersion strengthened ferritic steels are being considered for a number of advanced nuclear reactor applications because of their high strength and potential for high temperature application. Since these properties are attributed to the presence of a high density of very small (nanometer-sized) oxide clusters, there is interest in examining the radiation stability of such clusters. A novel experiment has been carried out to examine oxide nanocluster stability in a mechanically alloyed, oxide dispersion strengthened ferritic steel designated 12YWT. Pre-polished specimens were ion irradiated and the resulting microstructure was examined by atom probe tomography. After ion irradiation to ∼0.7 dpa with 150 keV Fe ions at 300 °C, a high number density of ∼4 nm-diameter nanoclusters was observed in the ferritic matrix. The nanoclusters are enriched in yttrium, titanium and oxygen, depleted in tungsten and chromium, and have a stoichiometry close to (Ti + Y):O. A similar cluster population was observed in the unirradiated materials, indicating that the ultrafine oxide nanoclusters are resistant to coarsening and dissolution under displacement cascade damage for the ion irradiation conditions used.     

The impact of the composition of structural steels on their corrosion stability in liquid Pb-Bi at 500 and 400 °C with different oxygen concentrations

In recent years, heavy liquid metals have found exercise as possible coolants and targets in the conversion of radioactive elements in accelerator driven systems (ADS). Liquid lead-bismuth eutectic alloy is one of candidates for this using tanks to its suitable nuclear and physical properties. Performed examination was aimed at research of compatibility choice materials for parts of ADS with liquid Pb-Bi eutectic alloy, influence of composition choice materials on their corrosion resistance, influence of temperature and oxygen content. We performed corrosion tests of 1000 h each on approximately 20 types of structural steels (austenitic, ferritic and martensitic) in convection loops with flowing Pb-Bi at 500 and 400 °C and using different oxygen concentrations. The impact of Fe, Cr, Ni, Mn, Si, Al and Mo content on the corrosion stability of these steels was measured without and after preliminary passivation through creating thin spinel or oxide layers on their surface.     

Sodium fast reactor evaluation: Core materials

In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.     

Creep behavior of oxide dispersion strengthened 8Cr-2WVTa and 8Cr-1W steels

Microstructures and creep behavior of two martensitic oxide dispersion strengthened (ODS) steels 8%Cr-2%W-0.2%V-0.1%Ta (J1) and 8%Cr-1%W (J2) with finely dispersed Y₂Ti₂O₇ have been investigated. Creep tests have been carried out at 670, 700 and 730 °C. Creep strength of J1 is stronger than that of any other ODS martensitic steels and the hoop strength of the ferritic ODS steel cladding. At the beginning of creep test, shrinkage was frequently observed for J1. This is one of the reasons for high creep strength of J1. The δ-ferrite, which is untransformed to austenite at hot isostatic press and hot rolling temperatures, was elongated along the rolling direction, and volume fraction of δ-ferrite in J1 is larger than J2. Although the elongated δ-ferrite affects the anisotropy of creep behavior, the extent of anisotropy in J1 is not so large as that of the ferritic ODS steel.     

Void-swelling in irons and ferritic steels: II. An experimental survey of materials irradiated in a fast reactor

The void-swelling response of a wide range of ferritic alloys irradiated in the Dounreay Fast Reactor to displacement doses up to 30 dpa (N/2) and covering the temperature range 380–615°C have been compared. The materials selected included high purity irons, together with commercial mild and low alloy steels, high chromium (12–14%) ferritic and martensitic stainless steels and a range of high purity binary iron-chromium alloys containing chromium contents up to 15%. The pure irons and binary iron-chromium alloys exhibited measurable but relatively low swellings (<1%) whilst all the commercial ferritic steels appeared to be void-swelling resistant, with swellings below the experimental detection limit (0.1%). The pattern emerging is thus one of overall swelling resistance in ferritic materials as a general class. Void-swelling in the pure iron peaked at two irradiation temperatures (~420, ~510°C), and the low magnitude of the swelling was rationalized in terms of the operation of solute-controlled swelling suppression mechanisms involving residual interstitial impurities. The complex functional dependence of peak-swelling on chromium content in the binary iron-chromium alloys was explained in terms of void-swelling suppression based on the presence of weak interactions between chromium atoms in solution and vacancies, modified by depletion of chromium from solid solution by α' precipitation at chromium contents exceeding 10%. The validation of the high swelling resistance of the 12% Cr martensitic stainless steels in a fast reactor environment provides confidence in the selection of these alloys as alternative core component materials for commercial fast reactor systems.     

A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can’t withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.     

Plastic deformation of ferritic grains in presence of ODS particles and irradiation-induced defect clusters: A 3D dislocation dynamics simulation study

Ferritic steels strengthened by oxide particle dispersions (ODS) are prime candidate for future nuclear applications. So far, the beneficial mechanical characteristics of ODS steels are not fully understood, in terms of dislocation-based mechanisms. In this work, three-dimensional discrete dislocation dynamics simulations were carried out to analyze pre and post-irradiation plastic deformation in ferritic grains, with and without ODS particles. In the absence of irradiation induced defect loops, ODS-grains are stronger and plastic strain is more localized than in the corresponding, particle-free grain. After irradiation however, ODS-grain become more resistant to loop-induced hardening, while plastic strain spreading is broader, with respect to particle-free grain. This effect is due to dislocations accumulating next to the precipitates, generating internal stress allowing cross-slipped dislocations to go past the irradiation induced defect loops. Cross-slip is therefore a key feature of our model, for explaining the beneficial role of ODS particles on post-irradiation plastic deformation.     

A systematic approach for the study of radiation-induced segregation/depletion at grain boundaries in steels

High Cr ferritic steels are candidate materials for structural applications in Gen-IV and fusion nuclear reactors. However, the relative contributions of irradiation conditions and materials microstructures on radiation-induced segregation or depletion of Cr at grain boundaries in ferritic steels are unclear. Here, the possibility of systematically analyzing the chemistry of the same grain boundary of known character during irradiation is demonstrated using a combination of electron back-scattered diffraction, atom-probe tomography and focused ion beam specimen preparation. This method provides a dynamic evolution of grain boundary chemistry as function of dose, spatial variations within the grain boundary plane, and quantification of minor solute elements such as carbon otherwise difficult to obtain experimentally.     

核电厂核级承压设备对铁素体材料的断裂韧性基本要求

本文概要介绍了铁素体材料构件的低温脆断的理论基础和抗脆断设计．总结并评价了ASME规范中对核电厂核安全级别承压设备铁素体材料抗脆断的断裂韧性要求．即核安全级别与材料的缺口冲击韧性要求之间存在相应的关系，从而保证了核电厂承压边界不会发生脆性破裂。     

Compatibility of FBR materials with sodium

In order to incorporate a procedure for the evaluation of the sodium environmental effects on core and structural materials into the elevated temperature structural design guide lines for fast breeder reactors, R&D on the sodium compatibility of the materials has been in progress in Japan Atomic Energy Agency. This paper reviews corrosion behavior in the sodium of conventional austenitic and ferritic steel. Simultaneously, the corrosion and mechanical properties of the materials for advanced FBRs, 12Cr steel and ODS steels are summarized, including the results of recent research.     

Irradiation creep of oxide dispersion strengthened (ODS) steels for advanced nuclear applications

Ferritic oxide dispersion strengthened steels with different microstructure were in-beam creep tested in a temperature range from 300 to 500 °C. Irradiation was by He-ions. Elongation was determined as a function of stress and irradiation damage rate. Damage was investigated by transmission electron microscopy. A thorough analysis of the loops developing during irradiation creep did not show any dependence of orientation or size on the direction of the applied stress. At 400 °C radiation induced segregation was found (most probably an iron aluminide) which had no effect on irradiation creep. No pronounced influence of microstructure or dispersoid size on the irradiation creep behavior was detected. Irradiation creep compliance of PM2000 with dispersoids of about 30 nm diameter were found to differ little from material with dispersoids of only 2-3 nm diameter. This is in contrast to thermal creep where dislocation-obstacle interactions are extremely important. An assessment of the technical relevance of irradiation creep in advanced nuclear systems is presented.