W.R. Johnson

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

Abstract This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr–2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54–2.53 dpa at 30–100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr–2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress–strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13–18% and the ductility by 8–36%. The effect of test temperature for the 9Cr–2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress–strain data predicted positive strain hardening during plastic instability.

Comparison of fission neutron and proton/spallation neutron irradiation effects on the tensile behavior of type 316 and 304 stainless steel

Abstract As part of the accelerator production of tritium and the spallation neutron source programs, the tensile properties of annealed 304L, 316LN and 316L stainless steel have been measured after proton and spallation neutron irradiation in the target region of the Los Alamos Neutron Science Center (LANSCE) accelerator (800 MeV, 1 mA) to a maximum dose of 12 dpa at temperatures ranging from 30 to 120 °C. In addition to the displacement damage produced from the irradiation, up to several thousand atomic parts per million (appm) of hydrogen and helium were produced in the irradiated material via spallation reactions. Results of tensile tests at temperatures from room temperature up to 164 °C show large increases in tensile yield strength, small increases in ultimate tensile strength, reductions in strain hardening capacity and reductions in ductility (uniform elongation and strain-to-necking) with increasing irradiation dose. A comparison of these data with the large database on tensile properties of type 316 stainless steel exposed to fission neutrons and tensile tested over the same temperature range show similar trends with regard to strength changes, but significantly larger reductions in ductility with irradiation dose were observed after irradiation in the spallation environment. The much higher amounts of helium and hydrogen produced through spallation in the LANSCE spectrum, compared to those developed in fission neutron irradiation environments at equivalent doses, may contribute to degradation in ductility.

Tensile and impact properties of V–4Cr–4Ti alloy heats 832665 and 832864 ☆

Abstract Two large heats of V–4Cr–4Ti alloy were produced in the US in the past few years. The first, 832665, was a 500 kg heat procured by the US Department of Energy for basic fusion structural materials research. The second, 832864, was a 1300 kg heat procured by General Atomics for the DIII-D radiative divertor upgrade. Both heats were produced by Oremet-Wah Chang (previously Teledyne Wah Chang of Albany). Tensile properties up to 800°C and Charpy V-notch impact properties down to liquid nitrogen temperature were measured for both heats. The product forms tested for both heats were rolled sheets annealed at 1000°C for 1 h in vacuum. Testing results show the behavior of the two heats to be similar and the reduction of strengths with temperature to be insignificant up to at least 750°C. Ductility of both materials is good in the test temperature range. Impact properties for both heats are excellent – no brittle failures at temperatures above −150°C. Compared to the data for previous smaller laboratory heats of 15–50 kg, the results show that scale-up of vanadium alloy ingot production to sizes useful for reactor blanket design can be successfully achieved as long as reasonable process control is implemented (H. Tsai, et al., Fusion Materials Semiannual Progress Report for Period Ending 30th June 1998, DOE/ER-0313/24, p. 3; H. Tsai, et al., Fusion Materials Semiannual Progress Report for Period Ending 31st December 1998, DOE/ER-0313/25, p. 3).

Tensile fracture characterization of braze joined copper-to-CFC coupon assemblies

Abstract A vacuum brazing process was used to join a broad spectrum of carbon-fiber reinforced carbon matrix composite (CFC) materials, machined into cylindrical coupons, between coupons of oxygen-free copper, the braze alloy was a copper-base alloy which contained only low activation elements (Al, Si, and Ti) relative to a titanium baseline specification. This demonstration was of particular importance for plasma facing components (PFCs) under design for use in the Tokamak Physics Experiment (TPX); the braze investigation was conducted by General Atomics for the Princeton Plasma Physics Laboratory. A tensile test of each brazed assembly was conducted. The results from the braze processing, testing, and fracture characterization studies of this reporting support the use of CFCs of varied fiber architecture and matrix processing in PFC designs for TPX. Further, the copper braze alloy investigated is now considered to be a viable candidate for a low-activation bond design. The prediction of plasma disruption-induced loads on the PFCs in TPX requires that joint strength between CFC tiles and their copper substrate be considered in design analysis and CFC selection.

Utilization of vanadium alloys in the DIII-D radiative divertor program

Abstract Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics in conjunction with Argonne National Laboratory and Oak Ridge National Laboratory has developed a plan for the utilization of vanadium alloys as part of the radiative divertor upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy. This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components and (2) demonstrating the in-service behavior of a vanadium alloy (V4Cr4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming radiative divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development efforts to support fabrication development and to resolve key issues related to environmental effects.

Metallurgical bonding development of V–4Cr–4Ti alloy for the DIII-D radiative divertor program

General Atomics (GA), in conjunction with the Department of Energy`s (DOE) DIII-D Program, is carrying out a plan to utilize a vanadium alloy in the DIII-D tokamak as part of the DIII-D Radiative Divertor (RD) upgrade. The V-4Cr-4Ti alloy has been selected in the U.S. as the leading candidate vanadium alloy for fusion applications. This alloy will be used for the divertor fabrication. Manufacturing development with the V-4Cr-4Ti alloy is a focus of the DIII-D RD Program. The RD structure, part of which will be fabricated from V-4Cr-4Ti alloy, will require many product forms and types of metal/metal bonded joints. Metallurgical bonding methods development on this vanadium alloy is therefore a key area of study by GA. Several solid state (non-fusion weld) and fusion weld joining methods are being investigated. To date, GA has been successful in producing ductile, high strength, vacuum leak tight joints by all of the methods under investigation. The solid state joining was accomplished in air, i.e., without the need for a vacuum or inert gas environment to prevent interstitial impurity contamination of the V-4Cr-4Ti alloy.

Vanadium alloys for the radiative divertor program of DIII-D

Vanadium alloys provide an attractive solution for fusion power plants as they exhibit a potential for low environmental impact due to low level of activation from neutron fluence and a relatively short half-life. They also have attractive material properties for use in a reactor. General Atomics along with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan to utilize vanadium alloys as part of the Radiative Divertor Project (RDP) modification for the DIII-D tokamak. The goal for using vanadium alloys is to provide a meaningful step towards developing advanced materials for fusion power applications by demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak in conjunction with developing essential fabrication technology for the manufacture of full-scale vanadium alloy components. A phased approach towards utilizing vanadium in DIII-D is being used starting with small coupons and samples, advancing to a small component, and finally a portion of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. A major portion of the program is research and development to support fabrication and resolve key issues related to environmental effects.

Engineering design of a radiative divertor for DIII-D

A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The radiative divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the radiative divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The radiative divertor design is very flexible, and will allow physics studies of the effects of slot width and length. The slot width is varied by installing graphite tiles of different geometry and can be accomplished in a shut down of less than 3 weeks. The change in slot length requires moving the structure vertically and could to be done in about 6-8 weeks. Slot lengths of 23, 33, and 43 cm have been chosen. Radiative divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D advanced divertor program form a foundation for the design work on the radiative divertor.

DESIGN, FABRICATION AND TESTING OF HELIUM-COOLED VANADIUM MODULE FOR FUSION APPLICATIONS

Vanadium alloys are attractive materials for fusion applications due to their low neutron activation and rapid decay of radioactivity with time. Design of high heat flux components with vanadium as the structural material is difficult due to its low thermal conductivity relative to copper and the lack of practical experience with fabrication of vanadium components. Similarly, helium is an attractive coolant for fusion power plants due to its chemical inertness, its transparency to neutrons, and stable heat transfer. However, there is a perceived difficulty that the use of helium as a coolant will limit the maximum heat flux on components. Reference 1 discusses the principle that heat transfer enhancement techniques reduce the pumping power for helium cooling, making it practical for cooling plasma facing components and General Atomics (GA) has demonstrated cooling of high heat flux components with helium coolant. A copper module designed by GA was successfully tested to a steady state heat flux level of 3200 W/cm 2 over small area and 1000 W/cm 2 over the entire 20 cm 2 area. As a continued effort to demonstrate practical application of fusion science, GA undertook the present effort to fabricate a vanadium module cooled with helium. Due to lower thermal conductivity of vanadium (6% of copper), this module will withstand about 300 W/cm 2 heat flux over the entire length. The module was fabricated from V-4Cr-4Ti alloy and is 228 mm long and 22.1 mm in diameter. The thickness of the vanadium tube is 1.76 mm. The internal flow path has been designed to enhance the heat transfer coefficient to a value of about 1 W/cm 2 −°C at a helium flow rate of 20 g/s. A thermal stress analysis of the design was performed to ensure that the stresses are within limits at a heat flux level of 300 W/cm 2 and a helium pressure of 4 MPa. The test module has been hydrostatically tested to 7 MPa pressure and helium leak checked. The module is ready to be tested at the helium loop (4 MPa pressure 20 g/s flow) at Sandia National Laboratory, Albuquerque. Future high heat flux testing is planned.

Performance of V–4Cr–4Ti material exposed to the DIII-D tokamak environment

A series of tests is being conducted in the DIII-D tokamak to determine the effects of environmental exposure on a V-4Cr-4Ti vanadium alloy. These tests are part of the effort to build and install a water-cooled vanadium alloy structure in the DIII-D radiative diverter upgrade. Data from the test series indicate that the performance of the V-4Cr4Ti alloy would not be significantly affected by environmental exposure. Interstitial absorption by the material appears to be limited to the surface, and neither the tensile nor the impact properties of the material appear to be affected by the exposure.