Candidate Alloys Research Articles

High heat load experiments for first wall materials

Plasma disruptions in tokamaks lead to high energy deposition for short times on the first wall. Melting and evaporation processes on the first wall decrease the structural integrity of this component because of reduction of thickness and initiation of cracks, leading to a shortening of the life time of the first wall. This experimental study is focussed on the melting, evaporation and resolidification behaviour dependence on energy density, using a neutral beam injector (beam energy: 120 keV) on a 10 MW test stand. Specimens tested were SUS 304, SUS 316. PCA (primary candidate alloy): Al, Cu. C and Mo-coated copper alloys, which are expected to be used for components. The pulse length was varied from 20 ms to 250 ms (1.6 MJ/m 2 to 20 MJ/m 2) at a constant heat flux of 80 MW/m. Metallographic analyses of the melted zone were carried out by SEM and XMA. Effects of one shot and multi shots on the melt layer thickness and evaporation loss were also studied.

Disclinations: The magnetism and the magnetic anisotropies of the rare-earth–3d transition-metal hard magnets (invited)

An important class of hard magnets, involving rare earth and 3d transition metals and sometimes metalloids, includes Nd2Fe14B and Nd2Fe17. We have noted a correlation between the local site magnetism in these two compounds and whether those sites lie on nets of so-called major ligand lines or disclinations, i.e., bond lines shared by six common nearest neighbors. We have proposed that a criterion for choosing candidate alloys with strong 3d moments is the occurrence of such disclination nets, and using this criterion, have listed several structures having 3d sites with this characteristic. We have also rationalized the crystal field anisotropies as relating to the orientation of the major ligand lines.

Effect of test temperature on tensile and fatigue properties of nickel-base heat-resistant alloys

A series of tensile and strain controlled low-cycle fatigue tests were conducted at temperatures ranging from RT to 900°C on a nickel-base heat-resistant alloy, Hastelloy XR-II, which is one of the candidate alloys for applications in the process heating high-temperature gas-cooled reactor (HTGR). Fatigue tests at room temperature and all tensile tests were conducted in air, while fatigue tests at and above 400°C were conducted in the simulated HTGR helium environment. In those tests the effect of test temperature on tensile and fatigue properties was investigated. The ductility minimum point was observed near 600° C, while tensile and fatigue strengths decreased with increasing test temperature. The fatigue lives estimated with the method proposed by Manson were compatible with the experimental results under the given conditions. For the specimens fatigued at and above 700°C, the percentage of the intergranular fracture mode gradually increased with increasing test temperature.

Development of a Coal Ash Corrosivity Index for High Temperature Corrosion

Current development of Advanced Steam Cycle coal-fired power plants requires superheater and reheater tubing alloys which can withstand severe conditions for high temperature corrosion. A corrosion equation to predict corrosion rates for candidate alloys has been developed by a study of deposits removed from steam generator tubes and from test probes installed in a boiler, supplemented by laboratory studies using synthetic coal ash. The corrosion equation predicts corrosion for a particular coal as a function of its content of sulfur, acid-soluble alkalies, and acid-soluble aklaline earths. Good agreement was obtained between the corrosion equation and 6000-hour tests using probes of TP347H and 17-14 CuMo.

High Damping Alloys—The Metallurgist's Cure for Unwanted Vibrations

AbstractHigh intrinsic damping in alloys finds many applications in the elimination of unwanted vibrations and reduction of acoustic noise. The increasing interest in these materials has resulted in the development of a new class of engineering alloys, the HIDAMETS (high-damping metals), specifically tailored for use in vibration-resistant structures. Numerous different physical mechanisms (usually controlled by the alloy microstructure) can be responsible for the level of internal friction or damping in a candidate alloy for a particular application. Consequently, alloy selection depends mainly upon the conditions to be encountered in service, i.e. temperature, frequency of vibration, amplitude of vibration, magnetic field, etc. Characterization of the internal friction in high damping alloys is described and some of the mechanisms involved are discussed. Results obtained for two different alloys are presented. The first is a commercial Mn-Cu alloy used for marine propellers and the second is a die-cast ...

Disclinations and magnetism in rare-earth–transition-metal hard magnets

An important class of magnets, involving rare-earth and 3d transition metals and sometimes metalloids, includes Nd2Fe14B and Nd2Fe17. We have noted a correlation between the local site magnetism in these two compounds and whether those sites lie on nets of so-called major ligand lines or disclinations, i.e., bond lines shared by six common nearest neighbors. We propose that a criterion for choosing candidate alloys with strong 3d moments is the occurrence of such disclination nets. We have developed a computer program for the calculation of these nets, and using this, several structures having 3d sites with this characteristic have been found.

The effect of 14 MeV neutron irradiation on the mechanical properties of Fe and stainless steels

The effect of 14 MeV neutron irradiation on the mechanical properties of pure Fe and fusion reactor candidate stainless steels such as JPCA (Japanese Prime Candidate Alloy) and JFMS (Japanese Ferritic/Martensitic Stainless Steel) have been investigated using small size tensile specimens. Both austenitic (JPCA) and ferritic/martensitic (JFMS) stainless steels showed significant hardening in the neutron fluence region between 1 sx 10 21 n/m 2 and 5 sx 10 22 n/m 2 for room temperatu irradiation and room temperature tensile testing. The increase of the yield stress roughly followed as 1 2 power of the neutron fluence. However, in lower fluence region behaviours were different, namely, no hardening was observed in JFMS and softening was observed in JPCA. The isochronal annealing experiments showed that the hardening recovers above 400°C. The results of pure Fe differed from the results of the alloys. The difference is attributed to the high mobility of radiation-induced defects in Fe.

Modification of the grain boundary microstructure of the austenitic PCA stainless steel to improve helium embrittlement resistance

Grain boundary MC precipitation was produced by a modified thermal-mechanical pretreatment in 25% cold-worked (CW) austenitic prime candidate alloy (PCA) stainless steel prior to HFIR irradiation. Postirradiation tensile results and fracture analysis showed that the modified material (B3) resisted helium embrittlement better than either solution annealed (SA) or 25% CW PCA irradiated at 500 to 600°C to ∼ 21 dpa and 1370 appm He. PCA SA and 25% CW were not embrittled at 300 to 400°C. Grain boundary MC survives in PCA-B3 during HFIR irradiation at 500°C but dissolves at 600°C; it does not form in either SA or 25% CW PCA during similar irradiation. The grain boundary MC appears to play an important role in the helium embrittlement resistance of PCA-B3.

A comparison of implantation-driven permeation characteristics of fusion reactor structural materials

Implantation-driven permeation experiments have been conducted on samples of the ferritic steel HT-9, the austenitic Primary Candidate Alloy (PCA) and the vanadium alloy V-15Cr-5Ti using D/sub 3//sup +/ ions under conditions that simulate charge-exchange neutral loading on a fusion reactor first wall. The steels all exhibited an initially intense permeation ''spike'' followed by an exponential decrease to low steady-state values. That spike was not evident in the V-15Cr-5Ti experiments. Steady-state permeation was highest in the vanadium alloy and lowest in the austenitic steel. Though permeation rates in the HT-9 were lower than those in V-15Cr-5Ti, permeation transients were much faster in HT-9 than in other materials tested. Ion-beam sputtering of the surface in the steel experiments resulted in enhanced remission at the front surface, whereas in the vanadium tests, recombination and diffusivity both appeared to diminish as the deuterium concentration rose. This may be due to a phase change in the material. We conclude that for conditions comparable to those of these experiments, tritium retention and loss in first wall structures made of steels will be less than in structures made of V-15Cr-5Ti.

Development of potential low activation ferritic and austenitic steels

New experimental steels were designed and tested as the reduced activation alternatives to the current austenitic and ferritic prime candidate alloys for applications to the first wall structural components of Tokamak reactors. The design was based on the experience and the knowledge implemented in the recent nuclear and non-nuclear material fields, in which optimization was attempted on the basic properties, keeping their levels equivalent to or better than those of the conventional candidates. Four proto-type alloys were melted for testing with the following compositional specifications: 0.1 C-8 Cr-2 W-0.2 V-Fe, 0.1 C-8 Cr-2 W-0.2 V-0.04 Ta-Fe, 0.25 C-25 Mn-14 Cr-0.2 Ti-Fe and 0.02 C-18 Ni-18 Cr-0.3 Ti-Fe. In the first phase of the evaluation program, tests were made to determine properties that included tensile, creep-rupture, impact before and after thermal aging, wet and dry corrosion, and weldability properties. These results are compaed with those of the conventionally known materials of the equivalent classes. Among the materials tested the ferritic alloy with 0.04 Ta showed particularly attractive features.

Program of United states-Japan collaborative testing in HFIR and ORR

The objective of the program of US-Japan collaborative testing in HFIR and ORR is to investigate the response of austenitic stainless steels to levels of neutron radiation damage anticipated in fusion reactors. Emphasis is placed on United States and Japanese Type 316 stainless steels and prime candidate alloys (PCA), but improved alloys are also included. The program consists of eight HFIR target capsules and two ORR spectral tailoring capsules with test temperatures of 60 to 600°C and fluences of 30 to 50 dpa. Helium contents after irradiation match or exceed those produced in a fusion reactor at the same dpa level. Types of data to be obtained include tensile, fatigue, crack growth, irradiation creep, swelling, and microstructural evolution. Five HFIR capsules have been irradiated, and testing is under way. Less than 1% swelling is observed for Type 316 stainless steel and PCA alloys irradiated to about 30 dpa at 300 to 500°C.

Fast nuclear methods of analysis in problems of reactor materials technology

The authors conduct a comprehensive review of nuclear reaction, tracer, radiochemical, and structural analytical techniques for assessing properties such as doping levels and effects, corrosion behavior and resistance, fuel-coolant and coolant-structure relationships, and physical radiation effects on a wide variety of candidate metals and alloys for nuclear and thermonuclear reactor materials. An extensive range of nuclear reactions for determining the diffusion and distribution of dopants and gas impurities is covered.

High-Pressure Creep Tests

The long-term effects of hydrogen at high temperature on the properties of alloys are relatively unknown. The most critical component in the engine is the heater head, which consists of the cylinders, tubings, and generator housing. Candidate alloys for these parts must meet property requirements and be of low cost for automotive application. This paper describes the design, development, and operation of a 20 MPa hydrogen high-temperature multispecimen creep-rupture apparatus possessing high sensitivity.

Behavior of pressurized tubes of candidate alloys in a simulated low-btu environment

Alloy tube specimens prepared from four candidate alloys for coal gasifiers (Alloy 800H, Type 310 stainless steel, HAYNES 188*, and IN-657) were exposed in an environment simulating low-Btu conditions at 815 °C. The test environment selected to represent slagging low-Btu gasifiers was an ambient pressure environment having oxygen and sulfur partial pressures of 1 x 10-14Pa (1 x 10-19 atm) and 1.3 x 10-2 Pa (1.3 x 10-7 atm), respectively. It was generated from a mixture of Ar, H2, H2O, and H2S. Lifetimes of these alloys under pressurization were compared with those observed in air and in other less sulfidizing simulated gasifier environments. The development of oxide/sulfide scales upon the outer surfaces was evaluated. Sulfidation was more severe with exposure to the low-Btu conditions compared to exposures in the other environments. The formation of external sulfides was stimulated by strain induced cracks. Alloy 800H experienced the greatest amount of sulfidation, which led to hot corrosion, i.e., the formation of molten eutectics. This contributed to a reduction in the endurance of this alloy. Although external sulfides developed on the other alloys, particularly at surface cracks, lifetimes were not noticeably affected for test durations of up to 400 hours. The development of these sulfides, however, reflects possible limitations of long-term performance under these conditions.

Chemo‐mechanical reliability of metallic materials in deep ocean environments — Part 1

The full development of marine technologies for the industrial exploitation of deep‐sea resources requires the availability of exhaustive engineering data on the degradation of constructional materials when immersed at great length in ocean environments. An overall review of behavioural figures from reliable sources allows to point out major weaknesses and uncertainties with candidate alloys and consequent demands for ameliorative and innovative investigation. Prominent objects of research can be envisaged, accordingly, in: (i) formulating low‐alloy steels whose surface is chemically convertible (by suitable pretreatments or free corrosion itself) so as to abate the rate of oxygen cathodic reduction i.e. the current density required for cathodic protection; (ii) singling out high‐strength steels whose resistance to corrosion‐fatigue and hydrogen embrittlement is in proportion to the strength level; (iii) developing cheaper alternatives to high‐molybdenum stainless alloys resistant to localised corrosion; (iii) combining (in optimal composites) the outstanding resistance of titanium to saltwater corrosion and the low‐cost good mechanical properties of steels.

極低温非磁性材料用溶着金属の機械的性質

Superconducting magnets in fusion reactor system produce enormous electromagnetic force which generates very high stress working on the containers of these magnets at cryogenic temperature. The candidate alloys for the structural material of containers are non magnetic stainless steels with improved yield strength and toughness at cryogenic temperature. Containers are produced as welded structures, so that these properties on the structural material are also indispensable for the weld metal.The chemical compositions of stainless steel weld metal are usually controlled to include a small quantity of δ-ferrite for the prevention of hot cracking in welds. The δ-ferrite is, however, ferromagnetic and causes the reduction in cryogenic toughness of weld metal. On the other hand, the weld metal of 19Cr-13Ni-5Mn type is known as fully austenitic one with very low susceptability on hot cracking in welds. The improvement of weld metal of this type resulted in the development of the fully austenitic weld metal of l7.5Cr-l6.5Ni-6.5Mn type with improvde cryogenic strengths. The effects of carbon and molybdenum contents on the tensile properties and Charpy impact properties were examined on this type of weld metal.Conclusions are as follows. It is possible to assure the yield strength of the weld metal of 2.3% molybdenum content higher than 1000 MPa when carbon content is higher than 0.07% and carbon plus nitrogen content is higher than 0.13%. While the weld metal decreases in the elongation and reduction in area as the addition of carbon, the weld metal which contains 0.13% of carbon and 2.3% of molybdenum indicates about 40% of elongation and about 300% of reduction in area at 4.2 K. The Charpy absorved energy and lateral expansion of the weld metal at 4.2 K are equal to those at 77 K. The weld metal decreases in the absorved energy and lateral expansion as the addition of carbon content, but the weld metal which contains 0.13% of carbon and 2.3% of molybdenum indicates about 60J of absorved energy and 0.8 mm of lateral expansion at 4.2 K. The relation between yield strength and absorbed energy at 4.2 K on the weld metal coincides with the lower limit line of absorbed energy of the stainless steels SUS304LN and SUS316LN of high yield strength.

Vanadium-base alloys for fusion reactor applications — a review

Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys with emphasis on the V-15Cr-5Ti alloy, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R&D) needs are defined. The relatively high thermal conductivity and low thermal expansion coefficient of vanadium-base alloys, which result in lower thermal stresses for a given heat flux compared to most other candidate alloys, should enhance the reactor wall-load and lifetime capability. Since the mechanical strength of vanadium-base alloys is retained at relatively high temperatures, higher operating temperatures are projected for these alloys than for austenitic or ferritic steels. The refractory metals, including vanadium, characteristically exhibit good corrosion resistance in purified liquid metals. The vanadium alloys also exhibit favorable neutronic properties which include lower parasitic neutron absorption leading to better tritium breeding performance, lower bulk nuclear heating rates, and lower helium generation rates compared to the steels. Limited data from moderate fluence neutron irradiations (fast breeder reactor) and high-damage level ion irradiations indicate good radiation damage resistance. Major concerns regarding the use of vanadium-base alloys for fusion reactor applications relate to cost, reactivity with oxygen/air, and the limited data base. The impact of these concerns and potential design solutions are discussed. Because of the corrosion/compatibility considerations, the vanadium alloys offer greater promise for liquid metal breeder concepts than for the solid breeder concepts.

Diffusion and Solubility of Tritium in Path a Prime Candidate Alloy

The tritium recoil injection technique was used to determine the tritium diffusion coefficient for Path A Prime Candidate Alloy (PCA) over the temperature range 323K to 523K. Dissolution tritium counting of samples saturated in a tritium-hydrogen mixture was used to determine the tritium solubility for PCA over the temperature range 673K to 973K. No evidence of high energy trapping due to the titanium carbide precipitates in PCA was seen.

High pressure hydrogen effect on creep-rupture properties of superalloys

The automotive Stirling engine, being developed by the United States Department of Energy and NASA, uses highpressure hydrogen as the working fluid. The longterm effects of hydrogen at high temperature on the properties of alloys are relatively unknown. The most critical component in the engine is the heater head, which consists of the cylinders, tubings, and regenerator housing. Candidate alloys for these parts must meet the property requirements and be of low cost for automotive application. Wrought alloys for tubing and cast alloys for cylinders and regenerator housings were tested in air and 15 MPa hydrogen at 650° to 925 °C. This paper evaluates the results for wrought alloy 19-9DL and cast alloy CRM-6D. Six specimens were evaluated in creeprupture simultaneously at 15 MPa H2 under constant initial stress conditions, and creep extensions were monitored with transducers placed internally. Air and H2 data were analyzed using the Orowan-Sherby-Dorn relationship. In both 19-9DL and CRM-6D, H2 did not significantly affect the stress level of 3500- hour rupture life, and the design limits of 28 MPa at 870 °C for 19-9DL tube alloy and 119 MPa at 775 °C for CRM-6D cast housing alloys were met. However, H2 had a deleterious effect on ductility in both alloys, and alloy development is under way.

Permeation of Deuterium Implanted into Fusion Reactor Materials

This paper reports on implantation-driven permeation experiments made for primary candidate alloy (PCA) and the ferritic steel HT-9 using deuterium ion beams from an accelerator. The results includ...