Prime Candidate Alloy Research Articles

Bend-fatigue properties of JPCA and Alloy800H specimens irradiated in a spallation environment

To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS) and spallation neutron source, post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) specimens has been carried out. The specimens tested in this study were made from the austenitic steel JPCA (Japan Primary Candidate Alloy) and high-Ni steel Alloy800H. The specimens were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as follows: the proton energy was 580MeV, irradiation temperatures ranged from 120 to 350°C and displacement damage levels ranged from 7.0 to 19.3dpa. Bend-fatigue tests were performed in air at room temperature under deflection control mode. The wave form of the control signal was a sine curve with a frequency of 26Hz. Fracture surface observation after the tests was done by SEM.The results on the irradiated JPCA in this study are identical to with the result of STIP-I specimens (−11dpa). Namely, the numbers of cycles to failure (Nf) were not changed by irradiation. Dpa dependence of Nf was not clearly seen in the irradiation conditions. In spallation environment, He atom production ratio is very high and most of He atoms are retained in the materials. In spite of large number of He atoms, all JPCA specimens show transgranular fracture surface.

Tensile mechanical properties of a stainless steel irradiated up to 19 dpa in the Swiss spallation neutron source

To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS), post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) specimens was carried out. The specimens tested in this study were made from the austenitic steel Japan primary candidate alloy (JPCA). The specimens were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as follows: the proton energy was 580MeV, irradiation temperatures ranged from 100 to 430°C, and displacement damage levels ranged from 7.1 to 19.5dpa. Tensile tests were performed in air at room temperature (RT), 250°C and 350°C. Fracture surface observation after the tests was done by Scanning electron microscope (SEM). Results of the tensile tests performed at R.T. showed the extra hardening of JPCA at higher dose compared to the fission neutron irradiated data. At the higher temperatures, 250°C and 350°C, the extra hardening was not observed. Degradation of ductility bottomed around 10dpa, and specimens kept their ductility until 19.5dpa. All specimens fractured in ductile manner.

The effect of temperature on the low cycle fatigue properties of a 15Cr-15Ni, Ti modified austenitic stainless steel

A titanium modified 15Cr-15Ni-2.5 Mo austenitic stainless steel designated as Alloy D9 is currently a favored material in liquid metal cooled fast breeder reactors for fuel clad and sub-assembly wrappers and is also being considered as a primary candidate alloy for fusion reactor first wall applications. The choice of this material in the nuclear industry is primarily based on its excellent resistance to radiation induced void swelling. In nuclear reactor applications the components are often subject to temperature gradient induced thermal stresses which are cyclic in nature as a result of start-ups, shutdowns and transients. Therefore there is a need for comprehensive understanding of both the macroscopic cyclic deformation behavior and the micromechanisms which influence such behavior. Further this understanding must include not only the behavior at maximum operating temperature but also the behavior at lower temperatures encountered during transients. Till date the cyclic stress response and fatigue life of Alloy D9 over a wide temperature range has not been examined and there is no information available on the deformation-induced microstructure during fatigue loadings. The purpose of this paper is to report some recent observations pertaining to the temperature dependence of cyclic stress response and evolving deformation substructure of Alloymore » D9 between 300 and 923 K.« less

Volatilization from PCA Steel Alloy

The mobilizations of key components from Primary Candidate Alloy (PCA) steel alloy have been measured with laboratory-scale experiments. The experiments indicate most of the mobilization from PCA steel is due to oxide formation and spalling but that the spalled particles are large enough to settle rapidly. Based on the experiments, models for the volatization of iron, manganese, and cobalt from PCA steel in steam and molybdenum from PCA steel in air have been derived.

Modelling of fracture process in annealed sheet of AISI 202 stainless steel

The similarities and differences in the cracking susceptibility between welding and resolidification after plasma disruption were examined experimentally using a number of primary candidate alloy samples with different chemical compositions. The product of the number density and the average depth of the cracks was measured after simulated disruption, employing a hydrogen ion beam as the heat source, and was compared with the Varestraint test result. An adequate correlation was observed between them, which indicates that the cracking susceptibility during plasma disruption can be well estimated from the welding cracking susceptibility.

Cracking susceptibility of stainless steel subjected to plasma disruption

The similarities and differences in the cracking susceptibility between welding and resolidification after plasma disruption were examined experimentally using a number of primary candidate alloy samples with different chemical compositions. The product of the number density and the average depth of the cracks was measured after simulated disruption, employing a hydrogen ion beam as the heat source, and was compared with the Varestraint test result. An adequate correlation was observed between them, which indicates that the cracking susceptibility during plasma disruption can be well estimated from the welding cracking susceptibility.

Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to ~ 178 dpa at ~ 400°C

The eighth and final irradiation segment for pressurized tubes constructed from the fusion Prime Candidate Alloy (PCA) has been completed in FFTF. At 178 dpa and ~ 400°C, the irradiation creep of 20% cold-worked PCA has become dominated by the “creep disappearance” phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels employed in this test. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however, requiring the use of several non-traditional techniques to extract the creep coefficients. No failures occurred in these tubes, even though the swelling ranged from ~ 20 to 40%.

Void swelling resistance of phosphorus-modified austenitic stainless steels during HFIR irradiation at 300–500°C to 57 dpa

The austenitic stainless steel prime candidate alloy (PCA) is a Ti-modified alloy (14Cr16Ni) developed for void swelling resistance at 400–600°C by the US Fusion Reactor Materials (FMR) program. During irradiation in the High Flux Isotope Reactor (HFIR) to 34 and 57 dpa (70–75 appm He/dpa) at 300–600°C, th e PCA (and the closely related D9 type alloy) has been shown to be highly susceptible to void swelling at 500°C. Generally such alloys show good void swelling resistance to higher doses during irradiation in fast-breeder reactors (FBRs, 0.5–1 appm He/dpa) in the 20–25% cold-worked (CW) condition. New multiply-stabilized, phosphorus-modified PCA alloys have been developed, with specific amounts and combinations of minor alloy elements added for improved MC formation and stability characteristics during irradiation. After HFIR irradiation at 300–500°C to 34–57 dpa, the multiply-stabilized, phosphorus-modified PCA and phosphorus-modified D9 alloy showed better swelling resistance (by density-change measurements) than similar alloys without phosphorus, particularly at 500°C. Microstructural studies after 34 dpa at 500°C showed that void swelling resistance in the multiply-stabilized, phosphorus-modified PCA alloys was directly related to the formation of ultrafine dispersions of MC precipitates during HFIR irradiation.

Irradiation Creep Behavior of Low Activation Steels in FFTF/MOTA

Irradiation creep behavior of low activation steels, developed as structural materials for fusion reactors, was investigated. The objective of this study is to provide a fundamental understanding of the irradiation creep mechanism based on microstructural evolution under fast neutron irradiation. Pressurized tube creep specimens fabricated from tube segments were irradiated in the Fast Flux Test Facility (FFTF), Materials Open Test Assembly (MOTA) during FFTF Cycles 11 and 12. This paper provides the first creep results obtained after FFTF cycle-11 irradiation. (2.25-3)Cr-(1-2)W bainitic steels and 12Cr-2W ferritic/martensitic steels showed equivalent or superior creep resistance to a modified 316 stainless steel, known as Japanese Prime Candidate Alloy (JPCA), under fast neutron irradiation up to 600 o C

Temperature dependence of the dislocation microstructure of PCA austenitic stainless steel irradiated in ORR spectrally-tailored experiments

Abstract Specimens of solution-annealed (SA) and 25% cold-worked (CW) prime-candidate-alloy (PCA) austenitic stainless steel were irradiated in ORR in spectrally-tailored experiments specially designed to produce fusion-relevant He/dpa ratios (12–18 appm He/dpa). SA and CW PCA were irradiated at 330 and 400°C to 13 dpa while only CW PCA was irradiated at 60, 200, 330 and 400°C to 7.4 dpa. Cavities and fine MC precipitates were only detectable at 330 and 400°C. Dislocations were a major component of the radiation-induced microstructure at 60–400°C. Mixtures of tiny “black-spot” loops, larger Frank loops, and network components of the total dislocation structure were very temperature dependent. Both SA and CW PCA contained Frank loops and network dislocations at 330 and 400°C, with SA PCA having more of both. Frank loop concentrations were maximum at 330°C and dislocations evolved most with dose at 400°C. At 60 and 200°C, the microstructure was dominated by very dense dispersions of tiny (1–3 nm diam) “black-spot” loops. No Frank loops were found at 60°C. Surprisingly, significant radiation-induced recovery of the as-cold-worked dislocation network occured in CW PCA at all temperatures. The nature of the radiation-induced microstructure makes a transition between 200 and 330°C.

Stress corrosion cracking and intergranular corrosion of neutron irradiated austenitic stainless steels

The effects of irradiation on stress corrosion cracking (SCC) and intergranular corrosion (IGC) susceptibility were investigated in solution-treated Fe19Cr9NiMn alloys and JPCA irradiated to 5.3×10 24 n/m 2 ( E > 1 MeV) at 573 K. In Fe19Cr9NiMn alloys, the irradiation enhanced IGC i n boiling HNO 3 + Cr 6+ solution when the alloys contained phosphorus and silicon and induced SCC in all the alloys with strain rate tensile tests in 571 K water containing 32 ppm oxygen. With increasing phosphorus and silicon contents. IGC was promoted but IGSCC was suppressed after irradiation. The results indicated that these elements are not the main contributors to irradiation-assisted SCC, although they affect SCC behavior. The Japanese Prime Candidate Alloy (JPCA) had better SCC resistance than Fe19Cr9NiMn alloys under the present irradiation condition.

Waste Disposal of Candidate Structural Materials in Fusion Reactors Utilizing Different Fuel Cycles

The management and disposal of the radioactive waste generated in any nuclear system are major safety and environmental concerns for the deployment of such a power source. In this paper the waste disposal rating is compared for four structural materials when used in deuterium-tritium, deuterium-deuterium, and D-{sup 3}He fusion reactors. The materials considered are HT-9, primary candidate alloy (PCA), Tenelon, and a modified HT-9. Generic models for the reactors are assumed such that each produces a fusion power of 10 MW/m of the axial length and has a sufficient shield/blanket to produce identical magnet damage rates. The latter is achieved by varying the material compositions and thicknesses. The results show that using the advanced fuel cycle D-{sup 3}He, with its low neutron yield, alleviates the activation problems and also allows considerable volume reduction of the radioactive waste. This cycle also permits the use of conventional alloys and at the same time satisfies the regulations criteria for shallow land burial of the low-level waste. In addition, and because of the low damage rate in the D-{sup 3}He reactors, the useful lifetimes of the materials are greatly increased.

Application of a Unified Constitutive Model to Finite Element Analysis of Notched Cylinder.

Rate dependence becomes more remarkable with increasing temperature in the inelastic behavior of engineering materials. When we deal with a stress/strain response in a structure, a sophisticated constitutive relationship must be applied to the stress analysis since the materials exhibit complicated characteristics under multiaxial stress states. The aim of this paper is to apply a unified constitutive model to finite element analysis of a notched cylinder. Experimental verification is carried out for a PCA material (Primary Candidate Alloy for a fusion reactor component). The local strain response near the notch root is measured by an electric-capacitance type strain gauge.

The feasibility of welding irradiated materials

Helium was implanted into solution-annealed (SA) 316 stainless steel, 20% cold-worked (CW) 316 stainless steel and titanium-modified Primary Candidate Alloy (PCA) through tritium decay to levels ranging from 0.18 to 256 appm. Full penetration welds were then made on helium-doped materials using gas tungsten arc welding (GTAW) under fully constrained conditions. Intergranular heat-affected zone (HAZ) cracking was observed in all of the materials containing greater than 1 appm He. Electron microscopy showed that the HAZ cracking originated from the growth and coalescence of grain boundary (GB) helium bubbles. Bubble growth kinetics in the HAZ is explained by stress-enhanced diffusive cavity growth. Results suggest that the propensity for HAZ cracking can be reduced by the pre-existing cold-worked structure and by finely-distributed MC precipitates that refine the distribution of helium bubbles and minimize the flow of vacancies in grain boundaries.

Study of near surface defects in He-implanted stainless steels by monoenergetic positron beam

Variable-energy (0–30 keV) positron beam studies have been carried out on 70 keV He-implanted fusion reactor candidate stainless steels, JPCA (Japanese Prime Candidate Alloy). Doppler broadening profiles of the positron annihilation as a function of incident positron energy were shown to be quite sensitive to defects introduced by He-implantation. A fitting model neglecting positron diffusion into the defects was used to analyze parameterized momentum data. The variation of the defect profile as a function of isochronal annealing temperature showed a clear correlation with the release rate of He measured by a quadrupole mass spectrometer.

Radiation damage of HFIR-irradiated candidate stainless steels for fusion applications

Data of the swelling behavior and tensile properties were summarized on two candidate alloys; Japanese Primary Candidate Alloy (JPCA) and Type 316 stainless steel (J316) irradiated in the High-Flux Isotope Reactor (HFIR) at temperatures ranging from 573 to 873 K to a maximum dose of 57 dpa (4200 appm He). Little temperature dependence of small swelling less than about 1% was observed in both solution-annealed (SA) and cold-worked (CW) JPCA and J316 at irradiation temperatures < 673 K up to nearly 56 dpa. At 773 K, the swelling increased rapidly in SA although CW alloys still had good swelling resistance. Tensile properties below 673 K were different from those obtained above 773 K; in the low temperature region, no substantial change of the work-hardening character was induced by irradiation, while in the high-temperature region, the apparent work-hardening exponent n and/or strength coefficient A became large. Through the observation of microstructural change, this is possibly due to precipitation. Substantial decrease in ductility was obtained in the high-temperature region with dose beyond 50 dpa.

Cross-section specimen preparation technique for nickel alloys and stainless steels

Ion-irradiation studies are often used to gain insights into the response of metals to intense neutron environments. Transmission electron microscopy allows the microstructural effects of the irradiation to be studied. When transverse specimen preparation techniques are employed, the entire irradiated region may be viewed in the electron microscope due to the limited range of the incident ions. One technique for the preparation of cross-section TEM specimens is presented. A reliable and repeatable procedure for electroplating nickel on nickel- and iron-based alloys has been developed. The plating bath electrolyte is a modified Wood nickel solution. The addition of a biological detergent to the electrolyte yields a non-porous plating layer. The application of a low plating current along with moderately intense agitation of the electrolyte results in a strong interfacial bond between the irradiated foil and the plated nickel. The strength of this interfacial bond determines the quality of the electron-transparent thin area that is obtainable. Successful plating has been achieved with ion-irradiated pure Fe-Ni-Cr ternary, nickel, Ni-Cu binaries, US Prime Candidate Alloy (PCA), and Soviet EP-838 stainless steel. An unirradiated low-activation ferritic steel has also been successfully plated. An electron-transparent thin area is achieved by two-step twin-jet electropolishing, supplemented by ion sputtering. Micrographs of several irradiated alloys displayed in cross-section using this technique are presented.

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.

Irradiation effects on the tensile behaviour of selected reference austenitic steels for fusion reactor applications

Abstract Tensile specimens of EC reference 316 stainless steel, the US reference 316 stainless steel, the US prime candidate alloy (PCA) and the Japanese prime candidate alloy were irradiated, after solution annealing or cold working, to displacement levels of up to 10 dpa in the Oak Ridge HFIR and in the Petten HFR reactors. Both the irradiated and the unirradiated specimens were tested at a strain rate of 4.6 × 10 −5 s −1 at the respective temperature of irradiation in the range 523–673 K. The results are analyzed to determine the role of chemistry, neutron spectrum, mechanical or thermal history and the temperature in modifying the deformation behaviour (yield strength, work hardening rate and ductility) of these materials due to irradiation. An attempt is also made to correlate these and other experimental observations on the deformation behaviour after irradiation to the microstructural changes expected due to irradiation.

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.