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General Kenneth Nichols and the Manhattan Project
Nichols
The Oak Ridger has published the latest in a series of articles about General Kenneth D. Nichols, the Manhattan Project, and the 1954 Atomic Energy Act. The series has been produced by Nichols’ grandniece Barbara Rogers Scollin and Oak Ridge (Tenn.) city historian David Ray Smith. Gen. Nichols (1907–2000) was the district engineer for the Manhattan Engineer District during the Manhattan Project.
As Smith and Scollin explain, Nichols “had supervision of the research and development connected with, and the design, construction, and operation of, all plants required to produce plutonium-239 and uranium-235, including the construction of the towns of Oak Ridge, Tennessee, and Richland, Washington. The responsibility of his position was massive as he oversaw a workforce of both military and civilian personnel of approximately 125,000; his Oak Ridge office became the center of the wartime atomic energy’s activities.”
I. N. Sviatoslavsky, M. E. Sawan, S. Majumdar
Fusion Science and Technology | Volume 44 | Number 1 | July 2003 | Pages 59-63
Technical Paper | Fusion Energy - MFE Chamber Technology | doi.org/10.13182/FST03-A310
Articles are hosted by Taylor and Francis Online.
The Advanced Power Extraction (APEX) program is exploring concepts for blanket designs that can enhance the potential of fusion while using both liquid and solid walls. In that context, an innovative blanket design of dry solid wall configuration, with potential for high neutron wall loading has been proposed. The blanket utilizes nanocomposited ferritic (NCF) steel structure (designated 12YWT), which has a maximum operating temperature of 800°C. The cooling/breeding material is Flibe (Li2BeF4), a low viscosity version of this molten salt, which has a melting temperature of 465°C and is compatible with ferritic steels up to 700°C. The dimensions for this study have been taken from ARIES-AT. The blanket module, which extends 0.3 m in the toroidal direction at mid-plane, is equipped with spiraling discs ramping from the bottom to the top. The coolant enters on the bottom at 500°C, then travels on the spiral discs, from the rear of the module to the front, then back to the rear, all the way to the top where it exits from the module at 590°C. On its way up, the coolant velocity is amplified at the first wall (FW) by centrifugal action, providing a high heat transfer coefficient for dissipating the high surface heating. The 3 mm thick FW is scalloped with semicircular projections inclined in the flow direction. This facilitates unimpeded smooth flow at the FW, while at the same time stiffening the FW against pressure, obviating the need for welded reinforcements. The discs are made of two halves assembled together with a Be pebble bed enclosed as a neutron multiplier. Preliminary neutronics analysis for the outboard blanket has shown a local tritium breeding ratio (TBR) of 1.36, and an energy multiplication (M) of 1.26 including contribution from a secondary module, using natural Li. The pressure drop of 0.56 MPa in the module produces a modified primary bending stress of 124 MPa where the allowable stress for this material at 750°C is 142 MPa. Without optimization, this blanket is capable of dissipating an average neutron wall loading of 6.4 MW/m2, with a peak value of 9.6 MW/m2 and a peak surface heating of 1.3 MW/m2. The coolant picks up energy in the secondary blanket exiting the reactor at 600°C. Assuming the use of a supercritical steam power cycle, an efficiency of 48% can be expected.
