Fusion Reactor Structures Research Articles

Effects of helium on irradiation response of reduced-activation ferritic-martensitic steels: Using nickel isotopes to simulate fusion neutron response

Understanding the effects of helium on microstructures and mechanical properties of reduced-activation ferritic-martensitic steels is important to use of these steels in fusion reactor structures. The 9Cr-2WVTa steels were doped with 58Ni and 60Ni isotopes at 2 weight percent to control the rate of transmutation helium generation. The samples were irradiated in the High Flux Isotope Reactor to ~24 displacements per atom at nominal temperatures of 300, 400, and 500°C, producing 228 and 7 atomic parts-per-million helium in the 58Ni- and 60Ni-doped samples, respectively. Transmission electron microscopy revealed a variety of precipitates and the radiation-induced dislocation loops and cavities (voids or helium bubbles). Tensile tests of the irradiated samples at the irradiation temperatures showed radiation-induced hardening at 300°C and radiation-induced softening at 400°C. Analysis indicates that the hardening primarily originated from the loops and cavities. The 58Ni-doped samples had greater strengthening contributions from loops and cavities, leading to higher hardening with lower ductility than the 60Ni-doped samples. The greater helium production of 58Ni did not show pronounced reductions in ductility of the samples.

A preliminary study on streaming effect through the gap of blanket modules for a K-DEMO fusion reactor

It is known that the blanket module in the fusion reactor is designed for slowing and absorbing neutrons, breeding tritium, and limiting radiation damage to other structures. These blankets are installed remotely at individual positions. In this process, gaps between the blanket modules can occur and the streaming effect through gaps has an effect on the performance of the fusion reactor. The streaming effect affects two aspects in the neutron transport problem: (1) activation of the fusion reactor structures and (2) radiation shielding. It is noted that the shape of the gap is locally different due to the specific shape of the fusion reactor. Also, the mechanical difference of the blanket modules changes the gap size. In this study, the streaming effect caused by different shape and size of the gap are evaluated for a K-DEMO fusion reactor. The blanket modules of the fusion reactor, which are the blanket, vacuum vessel, and water channel, were simply simulated. In each module, the neutron streaming effects were assessed by changing the gap size and shapes. The MCNP5 code was used for the evaluations with ENDF/B-VI cross-section library and SAB2002 thermal cross section library. It was noticed that the neutron flux was significantly changed due to the streaming effect. This study can give guidelines for the simplification of blanket modeling in radiation transport problems.

Alpha Emission Spectra of 27Al, 50,52Cr, 55Mn, 54,56Fe, 58,60Ni Nucleus for Neutron Induced Reaction

The design of novel nuclear facilities, fusion as well as fission reactors, requires the knowledge of all properties of relevant materials, including the nuclear differential cross sections for a careful selection. The nuclear cross sections data for gas production via particle (neutron, proton, alpha, etc.) induced reactions are great importance in the domain in the fusion reactor technology, particularly in the calculation of nuclear transmutation rates, nuclear heating and radiation damage due to gas formation. In fusion reactor structures, a serious damage mechanism has been gas production in the metallic resulting from diverse nuclear reactions, mainly through (n, p) and (n, α), (n, d), (n, t). In the present study, by using equilibrium reaction mechanisms, the (n, xα) reaction alpha emission spectra for 27Al, 50,52Cr, 55Mn, 54,56Fe, 58,60Ni isotopes were investigated from 9 to 15 MeV incident neutron energy. The equilibrium results have been calculated by using the hybrid model, the geometry dependent hybrid model. Calculation results have been also compared with the available measurements in literature.

Investigation of Some Structural Fusion Materials for (n, α) Reactions at the 14–15 MeV Energy Region

In fusion reactor structures, a serious damage mechanism has been gas production in the metallic resulting from diverse nuclear reactions, mainly through (n, α) and (n, p) reactions above a certain threshold energy. The neutron incident energy around 14–15 MeV is enough to excite the nucleus for the reactions such as (n, p), (n, d), (n, 2n), (n, t), and (n, α). Design of the fusion reactor, about the 14–15 MeV neutron incident energy reaction cross sections is of great importance for various target nuclei. In this study, the experimental data have been taken only at 14–15 MeV energy regions from EXFOR database. The (n, α) reactions for some structural fusion materials such as 27Al, 51V, 52Cr, 55Mn, 56Fe and 58Ni have calculated by using evaluated empirical formulas developed by Tel et al. at 14–15 MeV and calculated with the pre-equilibrium models up to 20 MeV. The calculated results are discussed and compared with the experimental data taken from EXFOR database.

Study on Some Structural Fusion Materials for (n,p) Reactions up to 30 MeV Energy

The neutron incident energy around 14–15 MeV is enough to excite the nucleus for the reactions such as (n,p), (n,d), (n,2n), (n,t) and (n,α). In fusion reactor structures, a serious damage mechanism has been gas production in the metallic resulting from diverse nuclear reactions, mainly through (n,p) and (n,α) and to some extent through (n,d) and (n,t) reactions above a certain threshold energy. So, working out the systematic of the neutron induced reaction cross sections is of great importance for the definition of the excitation function character for the given reaction taking place on various nuclei at different energies. In this study, (n,p) reactions for some structural fusion materials such as 27Al, 51V, 52Cr, 55Mn and 56Fe have been investigated. The new calculations on the excitation functions of 27Al(n,p)27Mg, 51V(n,p)51Ti, 52Cr(n,p)52V, 55Mn(n,p)55Cr and 56Fe(n,p)56Mn reactions have been carried out up to 30 MeV incident neutron energy. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. Also in the present work, the (n,p) reaction cross-sections have calculated by using evaluated empirical formulas developed by Tel et al. at 14–15 MeV energy. The calculated results are discussed and compared with the experimental data taken from EXFOR database.

Materials and design interface

The unprecedented demands faced by fusion structures primarily derive from severe time varying thermal-mechanical loading of complex, large scale, and highly interconnected heat transfer-energy conversion structures. This grand challenge is often much too narrowly couched in terms of the development of radiation damage resistant materials, while the enormously larger challenge is the creation of material systems and multifunctional structures. In addition, the fusion system designer is faced with the untenable situation that neither the fully functional materials, nor the requisite computational tools, nor experimental simulation facilities currently exist for reliable integrity and lifetime assessments of fusion reactor structures. Considering the absence of material information and design tools, neither the materials nor the fusion designer can follow standard design processes. The design process has to become actively materials-related while materials development must closely follow design process needs. This indispensible interaction between materials and design processes leads to a ‘concurrent materials-structure design’ path, which is necessary to meet the enormous materials-structural engineering challenges of fusion.

Utilization of Refractory Metals and Alloys in Fusion Reactor Structures

In design of fusion reactors, structural material selection is very crucial to improve reactor’s performance. Different types of materials have been proposed for use in fusion reactor structures. Among these materials, refractory metals and alloys having capability to withstand high temperatures and high neutron wall loads have been considered to get high power density in fusion reactors. However, these materials have insufficient technological database and are very expensive compared to steels. In addition to that, except chromium and some chromium alloys they show no low activation property. This study gives an overview of potential of refractory metals and alloys for possible use in fusion reactors.

Verification of Mechanical Properties of A V-4Cr-4Ti Alloy Using Finite Element Method

Vanadium alloys in the composition range around V-4Cr-4Ti have been proposed as candidate materials for fusion reactor applications and structures. These applications will require detailed characterization of constitutive and fracture properties. This study is aimed at understanding the basic constitutive and fracture mechanisms in vanadium alloys. Understanding of the basic constitutive and fracture mechanisms is achieved through a series of mechanical tests. These test results are combined with quantitative models of the underlying macro- and micromechanics. In addition to these experimental studies, finite element analysis (FEA) techniques are used to determine stress and strain fields to verify the constitutive law used in the fracture specimens.

Evaluation of fracture toughness by small-punch testing techniques using sharp notched specimens

The small punch (SP) testing method has been used to assess the reliability of industrial facilities such as fusion reactor structures and power-generation systems. Conventional SP tests have evaluated flow properties, transition temperature, fracture strain, and other mechanical properties by analyzing load-deflection curves. However, previous research has not used a fracture-mechanics approach that considers flaws, stress analysis and fracture toughness. In this study, in order to obtain the fracture toughness based on fracture mechanics by SP tests, a sharp notch was machined into the central part of the SP specimens. A stress-intensity factor for sharp-notched SP specimens is proposed from analysis of stress fields near the crack tip. The crack initiation point was determined by analysis of the load-deflection curves combined with acoustic emission signals. The fracture toughness of SA 508 Class 3 steel was successfully evaluated by using the load at the crack initiation point.

Advanced welding processes for fusion reactor fabrication

The huge scale and complexity of proposed fusion reactor structures has placed new demands upon fabrication technology and material performance. The need to weld substantial material thicknesses with severe service performance requirements, often in a multi-positional mode, over extended distances, certainly presents a formidable series of challenges. Of course, conventional welding processes can offer some solutions but welding times will be very long and component distortion potentially large. Fortunately, these challenges can be addressed, at least in part, by recent advances in electron beam and laser beam welding technology. Also recent developments in friction welding processes could offer benefits for the manufacture of some components. Specifically in the case of electron beam welding, TWI has developed a means of producing deep, narrow welds in a relatively poor vacuum. This new process variant called reduced pressure electron beam (RPEB) welding is being studied for fusion reactor vessel fabrication. Process advantages will be described. The work undertaken by TWI on behalf of the UKAEA for the EU Task No. T301/1 programme will be covered by a poster presentation at this conference. Recent advances in high power Nd:YAG laser welding and Friction Stir welding developments, which have taken, place at TWI will also be described.

Comprehensive Safety Analysis Code System for Nuclear Fusion Reactors IV: Preliminary Estimation of Dust Production Due to Plasma Disruptions

Dust production due to plasma disruptions has been investigated using a safety analysis code, which can calculate the plasma dynamics and thermal characteristics of fusion reactor structures simultaneously. We selected two fusion reactor designs in the International Thermonuclear Experimental Reactor (ITER ), i.e., the Engineering Design Activity (EDA) and the Conceptual Design Activity (CDA). The ITER/EDA will adopt beryllium for the plasma-facing component (PFC), and the ITER/CDA adopted graphite for PFC. The beryllium dust production in the ITER/EDA reactor will range from 7.0 to 10.3 kg/disruption, which strongly depends on vapor shield effects. The carbon dust production in the ITER/CDA reactor will range from 1.9 to 2.4 kg/disruption. However, the carbon dust will increase by as much as a factor of 2 to 5 because the effective latent heat of graphite has a large uncertainty under the extremely high heat flux during disruptions. For both, dust production from the first wall depends on the current quench time during disruptions. If the current quench time can be extended, the beryllium dust from the first wall will be minimized, and the carbon dust from there will be negligible.

Optimum alloy compositions in reduced-activation martensitic 9Cr steels for fusion reactor

In order to obtain potential reduced-activation ferritic steels suitable for fusion reactor structures, the effect of alloying elements W and V on the microstructural evolution, toughness, high-temperature creep and irradiation hardening behavior was investigated for simple 9Cr-W and 9Cr-V steels. The creep strength of the 9Cr-W steels increased but their toughness decreased with increasing W concentration. The 9Cr-V steels exhibited poor creep rupture strength, far below that of a conventional 9Cr-1MoVNb steel and poor toughness after aging at 873 K. It was also found that the Δ-ferrite should be avoided, because it degraded both the roughness and high-temperature creep strength. Based on the results on the simple steels, optimized martensitic 9Cr steels were alloy-designed from a standpoint of enough thoughness and high-temperature creep strength. Two kinds of optimized 9Cr steels with low and high levels of W were obtained; 9Cr-1WVTa and 9Cr-3WVTa. These steels indeed exhibited excellent toughness and creep strength, respectively. The 9Cr-1WVTa steel exhibiting an excellent roughness was shown to be the most promising for relatively low-temperature application below 500°C, where irradiation embrittlement is significant. The 9Cr-3WVTa steel was the most promising for high temperature application above 500°C from the standpoint of enough high-temperature strength.

Toward improved guidelines for reduced activation materials development in the US

A principal goal of the US fusion materials program is to develop new materials so as to minimize the activation of future fusion reactor structures. The objective is to improve the environmental acceptability of fusion power. The current guideline is that structural materials should meet the requirements of 10CFR61 for near-surface burial as radioactive waste. Analyses of the bases for the concentration limits in 10CFR61, and the trend toward conservatism in radioactive waste disposal, strongly suggest that the criteria in 10CFR61 will not be conservative when applied to fusion reactor waste materials. It is recommended that emphasis be placed on developing a strategy for the controlled recycling of radioactive materials within the nuclear industry. It is also recommended that a systematic sampling of the public perception of fusion power be initiated.

The effect of tungsten on dislocation recovery and precipitation behavior of low-activation martensitic 9Cr steels

The effect of W on dislocation recovery and precipitation behavior was investigated for martensitic 9Cr-(0,l,2,4)W-0.1C (wt pct) steels after quenching, tempering, and subsequent prolonged aging. The steels were low induced-radioactivation martensitic steels for fusion reactor structures, intended as a possible replacement for conventional (7 to 12)Cr-Mo steels. During tempering after quenching, homogeneous precipitation of fine W2C occurred in martensite, causing secondary hardening between 673 and 823 K. The softening above the secondary hardening temperature shifted to higher temperatures with increasing W concentration, which was correlated with the decrease in self-diffusion rates with increasing W concentration. Carbides M23C6 and M7C3 were precipitated in the 9Cr steel without W after high-temperature tempering at 1023 K. With increasing W concentration, M7C3 was replaced by M23C6, and M6C formed in addition to M23C6. During subsequent aging at temperatures between 823 and 973 K after tempering, the recovery of dislocations, the agglomeration of carbides, and the growth of martensite lath subgrains occurred. Intermetallic Fe2W Laves also precipitated in the δ-ferrite grains of the 9Cr-4W steel. The effect of W on dislocation recovery and precipitation behavior is discussed in detail.

Precipitation hardening owing to δ′ Al3Li in low activation Al–3 Li–12Si–3Mg alloy

The precipitation hardening behaviour owing to ordered δ′ Al3Li in an Al–3Li–12Si–3Mg (wt-%) based alloy was investigated during aging between room temperature and 523 K after quenching from between 723 and 823 K. The induced radioactivity level of this alloy is very low, suitable for fusion reactor structures. Effects of aging temperature, solutionising temperature, and pretreatments on precipitation hardening were examined in detail. Tensile properties were also examined in the as quenched condition as a function of testing temperature between room temperature and 673 K. Yield stress exhibited a positive temperature dependence between room temperature and 523 K and decreased abruptly above 523k with increasing test temperature. The precipitation behaviour of δ′ and the strengthening mechanisms owing to δ′ were discussed.MST/1100

Microstructural evolution in bainite, martensite, and δ ferrite of low activation Cr–2W ferritic steels

The microstructural evolution in (2–15)Cr–2W–0·1C (wt-%) firritic steels after quenching, tempering, and subsequent prolonged aging was investigated, using mainly transmission electron microscopy. The steels examined were low induced radioactivation ferritic steels for fusion reactor structures. With increasing Cr concentration, the matrix phase changed from bainite to martensite and a dual phase of martensite and δ ferrite. During tempering, homogeneous precipitation of fine W2C rich carbides occurred in bainite and martensite, causing secondary hardening between 673 and 823 K. With increasing tempering temperature, dislocation density decreased and carbides had a tendency to precipitate preferentially along interfaces such as bainite or martensite subgrain boundaries. During aging at high temperature, carbides increased in size and carbide reaction from W2C and M6C to stable M23C6 occurred. No carbide formed in δ ferrite. The precipitation sequence of carbides was analogous to that in conventional Cr–Mo steels.MST/1049

核融合炉用高Mnオーステナイト鋼の炭化物析出による粒界脆化

核融合炉用高Mnオーステナイト鋼の炭化物析出による粒界脆化

Discontinuous precipitation of σ-phase during recrystallisation in cold rolled Fe–10Cr–30Mn austenite

AbstractThe precipitation of σ-phase in the cold rolled specimens of an Fe–10Cr–30Mn steel was investigated by transmission electron microscopy after aging at 923 K. The specimens were prepared as a reduced activation austenitic steel for fusion reactor structures. During the aging, recrystallisation and discontinuous precipitation of σ-phase occurred simultaneously in the specimens with > 30% cold working (CW). The recrystallising interfaces moved toward the deformed and supersaturated solid solutions, leaving the precipitates of σ-phase behind the new recrystallised grains. The nucleation and growth behaviour of σ-phase is discussed.MST/779

Effect of specimen size on the ductile-brittle transition behavior and the fracture sequence of 9Cr-W steels

The effect of specimen size on the ductile-brittle transition behavior and the fracture sequence were investigated by means of Charpy absorbed energy measurement and fractography, using the full size, the half size and the one-third size V-notch specimens of 9Cr-W steels. The steels used are reduced-activation ferritic steels for fusion reactor structures. Attempts were made to correlate the impact data between the different specimen sizes by using normalizing parameters, such as nominal fracture area and nominal fracture volume for the upper shelf energy and ligament size for the ductile-brittle transition temperature. Fractography showed a similar fracture sequence for the three different sizes of the specimens.

Elemental Volatility of HT-9 Fusion Reactor Alloy

The volatility of elemental constituents from HT-9, a ferritic steel, proposed for fusion reactor structures, was investigated. Tests were conducted in flowing air at temperatures from 800 to 1200/sup 0/C for durations of 1 to 20 h. Elemental volatility was calculated in terms of the weight fraction of the element volatilized from the initial alloy; molybdenum, manganese, and nickel were the primary constituents volatilized. Comparisons with elemental volatilities observed for another candidate fusion reactor materials. Primary Candidate Alloy (PCA), an austenitic stainless steel, indicate significant differences between the volatilities of these steels that may impact fusion reactor safety analysis and alloy selection. Scanning electron microscopy and energy dispersive spectrometry were used to investigate the oxide layers formed on HT-9 and to measure elemental contents within these layers.