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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.”
A. Kemal Ziver, Sabu Shahdatullah, Matthew D. Eaton, Cassiano R. E. de Oliviera, Ron T. Ackroyd, Adrian P. Umpleby, Christopher C. Pain, Antony J. H. Goddard, James Fitzpatrick
Nuclear Technology | Volume 148 | Number 3 | December 2004 | Pages 223-234
Technical Paper | Fission Reactors | doi.org/10.13182/NT04-A3562
Articles are hosted by Taylor and Francis Online.
A homogenized whole-reactor cylindrical model of the Dounreay Fast Reactor has been constructed using both deterministic and Monte Carlo codes to determine neutron flux distributions inside the core and at various out-of-core components. The principal aim is to predict neutron-induced activation levels using both methods and make comparisons against the measured thermal reaction rates. Neutron transport calculations have been performed for a fixed source using a spatially lumped fission neutron distribution, which has been derived from measurements. The deterministic code used is based on the finite element approximation to the multigroup second-order even-parity neutron transport equation, which is implemented in the EVENT code. The Monte Carlo solutions were obtained using the MCNP4C code, in which neutron cross sections are represented in pointwise (or continuous) form. We have compared neutron spectra at various locations not only to show differences between using multigroup deterministic and continuous energy (point nuclear data) Monte Carlo methods but also to assess neutron-induced activation levels calculated using the spectra obtained from both methods. Results were also compared against experiments that were carried out to determine neutron-induced reaction rates. To determine activation levels, we employed the European Activation Code System FISPACT. We have found that the neutron spectra calculated at various in-core and out-of-core components show some differences, which mainly reflect the use of multigroup and point energy nuclear data libraries and methods employed, but these differences have not resulted in large errors on the calculated activation levels of materials that are important (such as steel components) for decommissioning studies of the reactor. The agreement of calculated reaction rates of thermal neutron detectors such as the 55Mn(n,)56Mn against measurements was satisfactory.