High temperature creep-fatigue design

Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowables and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed.     

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development. This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25–29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD.     

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development. This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25–29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD.     

Ferritic/martensitic steels for advanced nuclear reactors

Design concepts for the next generation of nuclear power reactors include water-cooled, gas-cooled, and liquid-metal-cooled reactors. Reactor conditions for several designs offer challenges for engineers and designers concerning which structural and cladding materials to use. Depending on operating conditions, some designs favor elevated-temperature ferritic/martensitic steels for in-core and out-of core applications. Such steels have been investigated in previous work on international fast reactor and fusion reactor research programs. Steels from these fission and fusion programs will provide reference materials for future fission applications. In addition, new elevated-temperature steels have been developed in recent years for conventional power systems that also need to be considered.     

Structural materials for fusion reactors

Fusion Reactors will require specially engineered structural materials, which will simultaneously satisfy the harsh conditions such as high thermo mechanical stresses, high heat loads and severe radiation damage without compromising on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics of structural materials. There exists considerable uncertainty in the nuclear data at such energies because there aren’t any strong enough sources for such neutrons except fusion reactors themselves! We thus encounter a problem of iterative nature in which we must try several experiments with the available materials in the near term. The development of such structural materials is thus going to require the experimental data of the kind that may be generated on reactors like ITER, high-performance modeling and a penetrating metallurgical insight to overcome technological challenges in terms of achieving required properties such as low activation by controlling the impurities, good thermo-mechanical properties by microstructure engineering, good chemical compatibility and high radiation resistance. These materials need to withstand a neutron wall load of the order of 2–3 MW/m², which can lead up to 30 dpa of radiation damage and 300 appm helium production per full power year in DEMO like reactors. Such conditions lead to unprecedented events related to the failure of materials due to irradiation creep, Ductile-Brittle Transition Temperature (DBTT) shift and helium embrittlement. The development of fusion materials program is oriented towards fulfilling the requirements of Test Blanket Modules, various prototype activities of SST-2 and DEMO reactor. The materials identified for first wall and blanket modules for Indian DEMO are LAFMS and ODS steels. The development program plan for these materials include (i) Manufacturing of LAFMS steel through VIM/VAR methods by controlling the impurities such as S, P and Si. (ii) ODS steel development with nano-size Y₂O₃ dispersoids in ferritic martensitic matrix by powder metallurgy route. The advanced structural materials like SiC_f /SiC composites and SiC_f /n-SiC are planned under National Fusion Program projects for indigenous development. An overview of the planned program in this direction will be presented.     

Thermal treatment improving creep properties of nitrogen-added Mod.9Cr-1Mo steels

Generation IV reactors are being developed to produce a reliable energy safely and with an economic benefit, because nuclear energy is being seriously considered to meet the increasing demand for a world-wide energy supply without environmental effects. Ferritic/martensitic steels are attracting attention as candidate materials for the Gen-IV reactors due to their high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. In recent years, new ferritic/martensitic steels have been developed for ultra supercritical fossil power plants through advanced technologies for steel fabrication. The microstructural stability of these materials for the pressure vessel, cladding and core structure of the VHTR and SFR is very important. Nitrogen is a precipitation hardening element, and the thermal stability of nitrides is superior to that of carbides. So the formation of nitrides may improve the thermal stability of the microstructure and eventually increase the creep rupture strength of high Cr steels. The effect of nitrogen on the creep rupture strength and microstructure evolution of nitrogen-added Mod.9Cr-1Mo steels has been studied. Creep testing was carried out at 873 and 923 K under constant load conditions. The optimum controlled Cr₂X precipitates were developed by special heat treatment, and they were not dissolved after a creep deformation. These fine and stable Cr₂X precipitates contributed to the increase of the creep rupture strength. The prior austenite grain size and martensite lath width were decreased by the resultant stable nitrides.     

A perspective on creep and fatigue issues in sodium cooled fast reactors

Development of advanced materials alongwith improved high temperature mechanical properties, particularly creep and fatigue are important and play a major role for the successful development of robust, safe and economical sodium cooled fast reactor (SFR) technology. The components of SFRs operate in demanding environments at high temperatures under complex creep, fatigue and creep-fatigue loading conditions. Based upon the service requirements in terms of different environments, temperature and loading conditions, different materials are chosen for different components. Ti modified 15Cr-15Ni austenitic stainless steel is chosen for clad and wrapper tubes in the reactor core, which experience high fast neutron flux of ～ 10¹⁵ ncm^?2s^?1 along with high temperatures. Type 316L(N) SS is used for out-of-core structural components like main and inner vessels, and sodium pipelines. For steam generators, modified 9Cr-1Mo steel is chosen for all the components, where liquid sodium and steam/water coexist. Some of the important experiences and exciting achievements in the areas of in-house materials development and its characterization in terms of creep, low cycle fatigue and creep-fatigue properties important to design of reactor components for core, out-of-core and steam generator applications are described in the paper. Future directions for materials research and development activities involving critical issues like radiation damage resistance along with improved mechanical properties for advanced clad and wrapper materials necessary for achieving high fuel burnup and design life up to 60 years for out-of-core structural components leading to economical nuclear energy have been highlighted.     

Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm³ were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 10⁴ hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.     

Creep rupture life prediction based on creep dissipation energy analysis

Structural materials used in sodium cooled fast reactors (SFRs) shall have good high temperature low cycle fatigue and creep properties, adequate weldability to fabricate large size components and shall be compatable with the liquid sodium environment in service. Austenitic stainless steels have been the natural choice for structural components of SFRs worldwide. The creep design life of SFR component is very long and is of the order of 40 years. This calls for robust creep life rediction models to convert short and medium term laboratory rupture data to design life. This paper discusses the application of creep dissipation energy concepts to predict creep rupture life of four nitrogen alloyed grades of 316 LN SS.     

用于超临界水堆燃料包壳的ODS铁素体钢的研究进展

 超临界水堆具有热效率高、系统简化、成本低等优点,成为第四代核反应堆中优先发展的堆型。ODS铁素体钢由于其优异的高温力学性能和良好的抗辐照能力成为超临界水堆包壳最有希望的候选材料。本文旨在回顾ODS铁素体钢制造工艺,包括机械合金化参数的优化,热处理工艺的选择以消除力学性能上的各向异性。根据超临界水堆包壳的服役条件,结合最新的实验数据,对ODS铁素体钢的高温力学性能、在超临界水中的耐腐蚀性以及中子辐照稳定性进行了总结和展望。     

Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at <0.3T _M (T _M is melting temperature) and up to 10 dpa (displacement per atom). Ferritic/martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.     

Materials requirements for fusion reactors

Once the physics of fusion devices is understood, which is expected to be achieved in the early 1980’s, one or more experimental power reactors (EPR) are planned which will pro-duce net electrical power. The structural material for the device will probably be a modi-fication of an austenitic stainless steel. Unlike fission reactors, whose pressure bound-aries are subjected to no or only light irradiation, the pressure boundary of a fusion reac-tor is subjected to high atomic displacement-damage and high production rates of trans-mutation products,e.g., helium and hydrogen. Hence, the design data base must include irradiated materials. Sincein situ testing to obtain tensile, fatigue, creep, crack-growth, stress-rupture, and swelling data is currently impossible for fusion reactor conditions, a program of service-temperature irradiations in fission reactors followed by postirradi-ation testing, simulation of fusion conditions, and low-fluence 14 MeV neutron-irradiation tests are planned. For the Demonstration Reactor (DEMO) expected to be built within ten years after the EPR, higher heat fluxes may require the use of refractory metals, at least for the first 20 cm. A partial data base may be provided by high-flux 14 MeV neutron sources being planned. Many materials other than those for structural components will be required in the EPR and DEMO. These include superconducting magnets, insulators, neutron reflectors and shields, and breeding materials. The rest of the device should utilize conventional materials except that portion involved in tritium confinement and re-covery. This paper is based on a presentation made at a symposium on “Materials Re-quirements for Unconventional Energy Systems” held at the Niagara Falls meeting of The Metallurgical Society of AIME, September 22, 1976, under the sponsorship of Non-Ferrous Metals and Ferrous Metals Committees.     

Type 316 austenitic steels for reactor vessel and internals in sodium fast reactors and their creep rupture properties

The type 316 stainless steels (316ss) are widely used as structural materials for reactor vessel and internals in sodium fast reactors (SFRs), which are agreed by designers of SFRs all over the world. Creep rupture properties become one of the important properties of the 316ss. The application of 316ss in fast reactors at abroad was briefly reviewed. Then the effect of alloying elements, microstructure and environmental media like neutron irradiation and sodium on creep rupture properties was summarized. More specifically, the alloying elements consist of carbon, nitrogen, molybdenum and phosphorus, while the microstructure consists of grain size and ferrite content. Finally, the current development status of 316ss for demonstration fast reactors in China was introduced. Furthermore, some suggestions were also proposed.     

抗辐照合金的发展与纳米结构ODS钢的抗辐照性能

简要介绍了抗辐照合金的发展,合金微观结构对合金抗辐照性能的影响,先进堆核心部件结构材料最佳的备选材料纳米结构氧化物弥散强化钢的特征性微观结构及其抗辐照性能.     

Continuum damage mechanics of constrained intergranular cavitation

A three-dimensional micromechanical model for polycrystalline materials undergoing constrained intergranular cavitation is proposed. The model combines advantages of phenomenological and micromechanical approaches, and can be calibrated from standard uniaxial creep data. When applied to creep analysis of two ferritic steels, the model accurately predicts the creep life under multiaxial stresses as well as the amount of creep damage.     

Relationships between phase stability and void swelling in Fe-Cr-Ni alloys during irradiation

Austenitic Fe-Cr-Ni steels are potential candidate alloys for structural materials in both fast breeder and magnetic fusion reactors. However, void swelling and phase instability during irradiation have been major problems limiting the properties and useful lifetimes of these materials and thus have been the subject of intensive investigations. Cavity nucleation in steels subjected to displacement damage is strongly influenced by the interactions between helium atoms and precipitates which are formed during irradiation. Several important mechanisms regarding gas atom-precipitate interactions and principles for the design of radiation-resistant alloys are critically examined. Central concepts derived from theory, including the critical radius and critical number of gas atoms, cavity-precipitate interactions, and relative sink strengths for point defects, are applied to the interpretation of experimental data. This paper is based on a presentation made in the “G. Marshall Pound Memorial Symposium on the Kinetics of Phase Transformations” presented as part of the 1990 fall meeting of TMS, October 8–12, 1990, in Detroit, MI, under the auspices of the ASM/MSD Phase Transformations Committee.     

Evolution of creep resistant 316 stainless steel for sodium cooled fast reactor applications

Sodium cooled fast reactors (SFRs) are designed to operate at high temperatures with an initial design life of about 40 years. Austenitic stainless steel (SS) types 304 and 316 and their variants have been generally used for out-of-core structural components of the reactor assembly system. The choice of these two grades of stainless steels is decided by several important factors such as high temperature mechanical properties like creep, low cycle fatigue and creep-fatigue interaction, compatibility with liquid sodium coolant, weldability, fabricability and cost. The components which operate in the creep temperature range are made of 316 SS. This material has been used extensively in the early SFRs. Studies on long term creep properties of 316 SS have clearly established the good creep resistance of this material and the microstructural stability at temperatures below 873 K. In view of the susceptibility of welded components to stress corrosion cracking, low carbon grades of 304 and 316 SS with alloying addition of nitrogen (designated as 304L(N) SS and 316L(N) SS) are used for structural components of later generation of SFRs. Nitrogen addition in the range of 0.06–0.08 wt% produces significant improvement in the creep properties of this material through solid solution strengthening and lowering of stacking fault energy. In view of the recent trends to increase the design life of SFRs to 60 years and more, it is necessary that non-replaceable structural components of reactor assembly have sufficient high temperature mechanical properties over such very long periods of operation. Increasing the nitrogen content from 0.06–0.08 wt % to levels of 0.12–0.14 wt% has been found to increase creep rupture life of 316LN SS by an order of magnitude. The beneficial effects of nitrogen are also extended to type 316 SS weld metal. This paper discusses the progressive improvements in the creep properties of 316 SS grade by varying the amounts of interstitial elements carbon and nitrogen.     

A Comparison of Creep Rupture Strength of Ferritic/Austenitic Dissimilar Weld Joints of Different Grades of Cr-Mo Ferritic Steels

Evaluations of creep rupture properties of dissimilar weld joints of 2.25Cr-1Mo, 9Cr-1Mo, and 9Cr-1MoVNb steels with Alloy 800 at 823 K were carried out. The joints were fabricated by a fusion welding process employing an INCONEL 182 weld electrode. All the joints displayed lower creep rupture strength than their respective ferritic steel base metals, and the strength reduction was greater in the 2.25Cr-1Mo steel joint and less in the 9Cr-1Mo steel joint. Failure location in the joints was found to shift from the ferritic steel base metal to the intercritical region of the heat-affected zone (HAZ) of the ferritic steel (type IV cracking) with the decrease in stress. At still lower stresses, the failure in the joints occurred at the ferritic/austenitic weld interface. The stress-life variation of the joints showed two-slope behavior and the slope change coincided with the occurrence of ferritic/austenitic weld interface cracking. Preferential creep cavitation in the soft intercritical HAZ induced type IV failure, whereas creep cavitation at the interfacial particles induced ferritic/austenitic weld interface cracking. Micromechanisms of the type IV failure and the ferritic/austenitic interface cracking in the dissimilar weld joint of the ferritic steels and relative cracking susceptibility of the joints are discussed based on microstructural investigation, mechanical testing, and finite element analysis (FEA) of the stress state across the joint.     

Review of small specimen test techniques for irradiation testing

Small specimen test technology has evolved out of the necessity to develop and monitor materials proposed for or used in nuclear power generation systems. Development of materials for improved cladding and in-core structures for fission reactors and assessment of core materials and pressure vessel steels already under irradiation necessitated the use of specimens which fit into existing irradiation space or which could be extracted from irradiated structures, such as cladding or ducts. Interest in simulating neutron irradiation by light and heavy ion irradiation led to the development of thin foil and wire geometry specimens. Further, interest in developing materials for fusion reactors has added additional constraints on specimen sizes associated with available irradiation volumes in existing and proposed high-energy neutron irradiation facilities. Consequently, a wide array of specimen geometries and test techniques has now been developed. It is the purpose of this paper to review these techniques and examine their status, problems, and potential for future applications. This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25–29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD.     

Review of small specimen test techniques for irradiation testing

Small specimen test technology has evolved out of the necessity to develop and monitor materials proposed for or used in nuclear power generation systems. Development of materials for improved cladding and in-core structures for fission reactors and assessment of core materials and pressure vessel steels already under irradiation necessitated the use of specimens which fit into existing irradiation space or which could be extracted from irradiated structures, such as cladding or ducts. Interest in simulating neutron irradiation by light and heavy ion irradiation led to the development of thin foil and wire geometry specimens. Further, interest in developing materials for fusion reactors has added additional constraints on specimen sizes associated with available irradiation volumes in existing and proposed high-energy neutron irradiation facilities. Consequently, a wide array of specimen geometries and test techniques has now been developed. It is the purpose of this paper to review these techniques and examine their status, problems, and potential for future applications. This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25–29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD.