Fusion Power Plant Components Research Articles

Fabrication of micro-scale fracture specimens for nuclear applications by direct laser writing

ABSTRACTThe structural integrity of nuclear fission and fusion power plant components is the focus of this research. The state of the art is using micro scale specimens milled with a focussed ion beam (FIB). Because of their very low volume such specimens can be lab tested, even when irradiated to low or medium level of activity. This offers a possibility of testing multiple specimens to investigate stochastic effects, e.g. effects of irradiation on the shift of the ductile to brittle transition. However, FIB milled specimens suffer from Ga contamination, to the degree that the validity of fracture data obtained on such specimens is questionable. We propose to use nano-additive manufacturing as an alternative to FIB for making micro scale fracture specimens. A combination of two-photon polymerization and electrodeposition and sputtering was used to manufacture micro-scale Brazilian disk fracture specimens (CBD), which are free from Ga and thus better suited for the study of irradiation effects on structural integrity. In this study Ni CBD specimens were made with 30 µm diameter and up to 13 µm thickness. The slot width varied between 1 µm to 2.9 µm width the corresponding slot length of between 7.5 µm and 8 µm. Consecutive FIB characterization shows that the specimens have polycrystalline microstructure with sub-µm grains. The work is ongoing making W CBD specimens and on reducing the slot width and using chemical vapor deposition fabrication.

Repeatability of irradiation damage and of recovery by post-irradiation annealing of EUROFER base steels

Former investigations clearly revealed that embrittlement and hardening of reduced-activation ferritic-martensitic (RAFM) steels after neutron irradiation can be reduced remarkably by post-irradiation annealing (PIA). This study demonstrates the repeatability of damage and recovery mechanisms for EUROFER-based and EUROFER-ODS-HIP-steels by using specimens of previous 330°C Bor 60 reactor (SSC RIAR) irradiations. Specimens were investigated in the irradiated state after a first PIA, after a subsequent further irradiation, and after a second annealing.Both tensile and Charpy-V impact tests clearly indicate that embrittlement occurs after the second irradiation of intermediately annealed (and, thus, recovered) material in the same order as after the first irradiation.In addition, a subsequent PIA leads to a complete recovery of the newly applied neutron damage, suggesting that lifetime of neutron-exposed fusion power plant components could be extended substantially by means of intermediate annealing.

Neutron-induced dpa, transmutations, gas production, and helium embrittlement of fusion materials

In a fusion reactor materials will be subjected to significant fluxes of high-energy neutrons. As well as causing radiation damage, the neutrons also initiate nuclear reactions leading to changes in the chemical composition of materials (transmutation). Many of these reactions produce gases, particularly helium, which cause additional swelling and embrittlement of materials. This paper investigates, using a combination of neutron-transport and inventory calculations, the variation in displacements per atom (dpa) and helium production levels as a function of position within the high flux regions of a recent conceptual model for the ‘next-step’ fusion device DEMO. Subsequently, the gas production rates are used to provide revised estimates, based on new density-functional-theory results, for the critical component lifetimes associated with the helium-induced grain-boundary embrittlement of materials. The revised estimates give more optimistic projections for the lifetimes of materials in a fusion power plant compared to a previous study, while at the same time indicating that helium embrittlement remains one of the most significant factors controlling the structural integrity of fusion power plant components.

Rhenium and molybdenum coatings for dendritic tungsten armors of plasma facing components: Concept, problems, and solutions

Abstract Fusion technology development requires materials resistant to heat, erosion and activation. Tungsten (W) could fulfill these requirements for plasma facing components of fusion power plants. A method developed by SNL to enhance the thermal and dimensional stability of W requires depositing a thin coating on the surface of dendritic W armors for IFE plasma facing components. This paper examines the waste disposal rating (WDR) associated with the activation of thin coatings of Re and Mo on the ARIES-ACT W-based divertor. Our results indicate that Re coatings must be limited to fewer than 30 μm in order to classify the Re/W-based divertor as low-level waste (LLW). Mo coatings that are larger than a few microns thick exceed the LLW disposal limit. It is recommended that, in addition to limiting the Mo thickness to 25 μm, natural Mo should be tailored to remove the Mo-94, Mo-98, and Mo-100 isotopes in order to reduce the WDR of the Mo/W-based divertor below the LLW disposal limit. If the isotopic tailoring process is inefficient or costly, SNL could focus their effort on the further development of the Re dendrite coating while Mo experiments would be regarded as a comparative study only.

Tensile and fracture toughness properties of the nanostructured oxide dispersion strengthened ferritic alloy 13Cr–1W–0.3Ti–0.3Y2O3

The realization of fusion power as an attractive energy source requires advanced structural materials that can cope with ultra-severe thermo-mechanical loads and high neutron fluxes experienced by fusion power plant components, such as the first wall, divertor and blanket structures. Towards this end, two variants of a 13Cr–1W–0.3Ti–0.3Y2O3 reduced activation ferritic (RAF-) ODS steel were produced by ball milling phase blended Fe–13Cr–1W, 0.3Y203 and 0.3Ti powders in both argon and hydrogen atmospheres. The milled powders were consolidated by hot isostatic pressing (HIP). The as-HIPed alloys were then hot rolled into 6mm plates. Microstructural, tensile and fracture toughness characterization of the hot rolled alloys are summarized here and compared to results previously reported for the as-HIPed condition.

Design issues and implications for structural integrity of fusion power plant components

The criteria and implications for successful design, licensing and power plant operation are assessed, and imposed constraints and limitations are examined. The design of a reliable fusion power plant is dependent on the availability of licensed nuclear materials and the structural–thermal loading conditions during normal and abnormal events. Various conditions in a tokamak lead to structural damage and possible failure. Taking into consideration all the possible structural failure mechanisms, the most likely are combinations of fatigue and creep. Issues encountered in the fusion environment are the significant amount of irradiation creep, the large ratio of helium production to displacement damage, and the degradation of fatigue strength and ductility, effects which are even encountered at low temperatures. Design codes distinguish between failure criteria under steady and transient loads, and lay down rules for failure prediction under combined creep–fatigue conditions. Currently, there are no established fusion specific licensing processes or component design codes. Any limits imposed on designs or performance are taken from existing design codes developed by the fission industry. There is a need to initiate the process of defining and developing tools for the design and licensing of fusion components and facilities to ensure nuclear safety.

Clearance considerations for slightly-irradiated components of fusion power plants

The development of commercial fusion plants includes the demonstration that the waste burden for future generations would be avoided. In order to minimize the radioactive material requiring long-term storage, maximum emphasis should be placed on recycling and clearance, avoiding geologic disposal. Clearance is the most desirable option for components like the vessel, shield, magnets and bioshield, which make up a great part of the mass and are only mildly irradiated. Recently, the International Atomic Energy Agency, the US Nuclear Regulatory Commission and other institutions have revised clearance guidelines for nuclear applications. In this paper, the implications of these new standards, particularly for slightly irradiated fusion materials, are considered. The amount of clearable materials is lower and/or the cooling period is longer than previously estimated. It becomes more evident now that improvements to the clearance data are needed to include many missing fusion-specific nuclides, and that the control and measurement of impurity levels, even for materials subjected to low-neutron exposure, is important. Segregation of components into more basic parts also plays a key role in minimizing the amount of waste. Finally, the issue of public acceptability of cleared materials is discussed since the unrestricted release of radioactive materials to the market is still controversial, despite the extremely low level of activity (⩽10 µSv/y)—way below the natural background.

Elastic–plastic FEM analysis on low cycle fatigue behavior for alumina dispersion-strengthened copper/stainless steel joint

Since the first wall and divertor components of fusion power plants are subjected to severe stresses caused by thermal expansion and electromagnetic forces, it is important to evaluate the fatigue strength of joints. In this study, elastic–plastic finite element analysis was performed for low cycle fatigue behavior of stainless steel/alumina dispersion-strengthened copper (DS Cu) joint in order to investigate the fatigue life and the fracture behavior of the joint. The results showed that a strain concentration occurred at the interface during low cycle fatigue, but as the strain range increased the strain concentration shifted away from the interface and into the DS Cu. The fatigue life and fracture location were evaluated taking into account of the strain concentration. Predictions of the fatigue life and fracture location were consistent with those measured by the low cycle fatigue test.

Range of blanket concepts from near term solutions to advanced concepts

Breeding blankets are key components of fusion power plants and determine to a large degree their attractiveness. There is a large range of possible blanket concepts, characterized by breeding material, structural material, coolant, and geometrical arrangement. On one extreme side there are concepts requiring only a modest extrapolation of the present technology, but with limited attractiveness. On the other side there are concepts with the potential for very attractive plants, but involving a considerable development risk. Therefore, the selection of blanket concepts depends on the overall strategy for fusion power plant development. To ensure sufficient support over the long development time, it has to be shown in a credible way, that (a) there are feasible blanket concepts for fusion power plants which can be developed with a high confidence of success, and (b) the final product will be an attractive power plant which can be operated safely, with minimum impact on the environment, and at an acceptable cost of electricity. This requires an evaluation of the entire range of blanket concepts, starting from near term solutions and reaching up to very advanced designs with a potential for exceptionally high performance. Typical examples of candidate blanket concepts are compared in this study. The blanket development strategy developed in the EU is described as an example.

Structural materials requirements for in-vessel components of fusion power plants

Structural materials requirements for in-vessel components of fusion power plants

Structural analysis and design against failure of fusion power plant components with special reference to pulsed operation

This study presents the theory of structural analysis and design code rules and constraints as applied to predict lifetime and prevent failure of components and is tailored to deal with the specific design requirements of fusion components. Example calculations are performed for a proposed ITER-like first wall using finite element analysis codes and an assessment of the structural behaviour is also included. The results of the calculations are used to arrive at a proposed design methodology, which accounts for the structural mechanisms likely to occur when operating a tokamak in pulsed mode, as proposed for ITER.