Gordon Willcutt

The mechanical properties of 316L/304L stainless steels, Alloy 718 and Mod 9Cr-1Mo after irradiation in a spallation environment

Abstract The Accelerator Production of Tritium (APT) project proposes to use a 1.0 GeV, 100 mA proton beam to produce neutrons via spallation reactions in a tungsten target. The neutrons are multiplied and moderated in a lead/aluminum/water blanket and then captured in 3 He to form tritium. The materials in the target and blanket region are exposed to protons and neutrons with energies into the GeV range. The effect of irradiation on the tensile and fracture toughness properties of candidate APT materials, 316L and 304L stainless steel (annealed), modified (Mod) 9Cr–1Mo steel, and Alloy 718 (precipitation hardened), was measured on tensile and fracture toughness specimens irradiated at the Los Alamos Neutron Science Center accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The irradiation temperatures ranged from 50°C to 164°C, prototypic of those expected in the APT target/blanket. The maximum achieved proton fluence was 4.5×10 21 p / cm 2 for the materials in the center of the beam. This maximum exposure translates to a dpa of 12 and the generation of 10 000 appm H and 1000 appm He for the Type 304L stainless steel tensile specimens. Specimens were tested at the irradiation temperature of 50–164°C. Less than 1 dpa of exposure reduced the uniform elongation of the Alloy 718 (precipitation hardened) and Mod 9Cr–1Mo to less than 2%. This same dose reduced the fracture toughness by 50%. Approximately 4 dpa of exposure was required to reduce the uniform elongation of the austenitic stainless steels (304L and 316L) to less than 2%. The yield stress of the austenitic steels increased to more than twice its non-irradiated value after less than 1 dpa. The fracture toughness reduced significantly by 4 dpa to ∼100 MPa m1/2. These results are discussed and compared with results of similar materials irradiated in fission reactor environments.

Performance of a Clad Tungsten Rod Spallation Neutron Source Target

Abstract Tungsten rods, slip-clad with Type 304L stainless steel, performed successfully as a spallation neutron source target operating to a peak fluence of ~4 × 1021 p/cm2. The target was used as a neutron source during the Accelerator Production of Tritium (APT) materials irradiation program at the Los Alamos Neutron Science Center. Tungsten rods of 2.642-mm diameter were slip-fit in Type 304L stainless steel tubes that had an inner diameter of 2.667 mm. The radial gap was filled with helium at atmospheric pressure and room temperature. Los Alamos High Energy Transport (LAHET) calculations suggest a time-averaged peak power deposition in the W of 2.25 kW/cm3. Thermal-hydraulic calculations indicate that the peak centerline W temperature reached 271°C. The LAHET calculations were also used to predict neutron and proton fluxes and spectra for the complex geometry used in the irradiation program. Activation foil sets distributed throughout the experiment were used to determine target neutronics performance as a comparison to the LAHET calculations. Examination of the irradiated target assemblies revealed no significant surface degradation or corrosion on either the Type 304L or the W surfaces. However, it was clear that the irradiation changed material properties because post-proton-irradiation measurements on Type 304L test samples from the APT program demonstrated increases in the yield strength and decreases in the ductility and fracture toughness with increasing dose, and the wrought W rod samples became brittle. Fortunately, the slip-clad target design subjects the materials to very low stress.

Fracture Toughness Characterization of 304L and 316L Austenitic Stainless Steels and Alloy 718 After Irradiation in High-Energy, Mixed Proton/Neutron Spectrum

This paper describes the fracture toughness characterization of annealed 304L and 316L stainless steels and precipitation hardened Alloy 718, performed at the Oak Ridge National Laboratory as a part of the experimental design and development for the Accelerator Production of Tritium (APT) target/blanket system. Materials were irradiated at 25 to 200C by high-energy protons and neutrons from an 800-MeV, 1-mA proton beam at the Los Alamos Neutron Science Center (LANSCE). The proton flux produced in LANSCE is nearly prototypic of anticipated conditions for significant portions of the APT target/blanket system. The objective of this testing program was to determine the change in crack-extension resistance of candidate APT materials from irradiation at prototypic APT temperatures and proton and neutron fluxes. J-integralresistance (J-R) curve toughness tests were conducted in general accordance with the American Society for Testing and Materials Standard Test Method for Measurement of Fracture Toughness, E 1820-99, with a computer-controlled test and data acquisition system. J-R curves were obtained from subsize disk-shaped compact tension specimens (12.5 mm in diameter) with thicknesses of 4 mm or 2 mm. Irradiation up to 12 dpa significantly reduced the fracture toughness of these materials. Alloy 718 had the lowest fracture toughness in both the unirradiated and irradiated conditions. All irradiated specimens of Alloy 718 failed by sudden unstable crack extension regardless of dose or test temperature. Type 304L and 316L stainless steels had a high level of fracture toughness in the unirradiated condition and exhibited reduction in fracture toughness to saturation levels of 65 to 100 MPa/m. The present reduction in fracture toughness is similar to changes reported from fission reactor studies. However, the currently observed losses in toughness appear to saturate at doses slightly lower than the dose required for saturation in reactor-irradiated steels. This difference might be attributed to the increased helium and hydrogen production associated with irradiation in the high-energy, mixed proton/neutron spectrum.

Thermal-Hydraulic Design of the Accelerator Production of Tritium Tungsten Neutron Source

The thermal-hydraulic design of the accelerator production of tritium (APT) tungsten neutron source is presented. A carefully engineered thermal-hydraulic design is required to remove the deposited power effectively during normal operations and remove the decay power during plant shutdown and postulated accidents. For steady-state operations and operational and anticipated transients, the design criterion is to maintain single-phase flow conditions with a margin to onset of nucleate boiling. The margin is determined based on phenomenological and geometric uncertainties associated with the design. A large margin to thermal excursion limits, such as critical heat flux and onset of flow instability, also is maintained during normal operations. In general, a very robust thermal-hydraulic design can be accomplished using the traditional models and correlations available in the engineering literature. However, two issues require further attention: maintaining adequate flows in a parallel network of flow channels and minimizing the volume fraction of heavy water to maximize tritium production. The design uses ladderlike structures that contain clad tungsten cylinders in the rungs that have coolant supplied and removed by the vertical ladder rails. Because the power density drops in the beam direction, the thickness of the tungsten cylinders is increased with increasing beam penetration length. The cooling requirement is determined using a conservative criterion where the minimum wall subcooling inside the rungs is at least 40°C and the minimum Reynolds number is 6000. Initial flow distribution tests were conducted with a full-scale model of an APT ladder assembly based on a preliminary design. Flow distributions can be made more even by using a larger riser than downcomer and also by increasing the flow resistance across each rung. The calculations discussed assume nominal dimensions, even though the power deposition and removal use a conservative approach. The effect of manufacturing tolerances will be investigated in future research. Also, the applicability of the critical heat flux and onset of flow instability models to small coolant channels is being verified experimentally. Further design optimization will be possible when these studies are completed.

Alert-derivative bimodal space power and propulsion systems

Safe, reliable, low‐mass bimodal space power and propulsion systems could have numerous civilian and military applications. This paper discusses potential bimodal systems that could be derived from the ALERT space fission power supply concept (Ranken 1990). These bimodal concepts have the potential for providing 5 to 10 kW of electrical power and a total impulse of 100 MN‐s at an average specific impulse of 770 s. System mass is on the order of 1000 kg.