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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.”
Yasunori Bessho, Yuichiro Yoshimoto, Osamu Yokomizo, Ryutaro Yamashita, Masumi Ishikawa, Akio Toba
Nuclear Technology | Volume 117 | Number 3 | March 1997 | Pages 281-292
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT97-A35342
Articles are hosted by Taylor and Francis Online.
Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates cal-culational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and a control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bun dle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the onedimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module. In the upper plenum/separator module, the upper plenum is modeled as a single volume in the thermal-equilibrium state and water spiraling in the separator is modeled by an effective length in the momentum equation. In the feedwater/recirculation flow module, the single-phase flow model is solved with the assumption of incompressive flow. Finally, the control system module includes the recirculation flow control minimodule, the pressure control minimodule, and the feedwater control minimodule, as well as the interlock functions, which work during a transient to allow analysis of general transient phenomena. The program was verified to provide satisfactory results within reasonable computational time based on application analysis of stability and scram phenomena in a BWR-5 type plant.
