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A MONTE CARLO CODE FOR THE TRANSPORT OF NEUTRONS

Description: A Monte Carlo code for the BM 704 computer was written to study the transport of neutrons in a uniform heterogeneous lattice of cylindrical fuel assemblies. Models of the geometric and physical processes are used to obtain the neutron age and migration area; the flux as a function of position, energy, and direction; and absorption data from which thermal utilization and the multiplication constant k may be calculated. (auth)
Date: December 1, 1959
Creator: Baxter, W.V.
Partner: UNT Libraries Government Documents Department

Preliminary Operationai Hazards Summary Report for the Task 2 Thermoelectric Generator

Description: The operational hazards associated with the use of an isotope-fueled auxiliary power unit for a satellite mission are described. The effects of missile about on the generator are discussed. The generator design is described, and the properties of the various fuel forms are investigated. The characteristics of the fuel capsules and the provisions for biological shielding are also described. Integration of the generator into a typical missile system is discussed. Hazards and procedures of transporting and handling the fuel cores from fabrication to launchlng are considered. Aborted missions are defined, and the forces acting on the generator during abort are described. (W.D.M.)
Date: December 1, 1959
Creator: Dix, G. P., Jr.
Partner: UNT Libraries Government Documents Department

Analysis of Radiation From Hnpf Cold Traps and Primary Sodium Pumps During Removal and Shipping

Description: The expected maximum contamination of the HNPF cold traps and primary sodium pumps was determined along with the maximum dose rates from these components during removal and shipping. Suitable shielding for casks to be used in the removal operation and for shipping these components away from the reactor site is specified. Access to an unshielded cold trap is limited by high dose rates, i.e., 100 mr/hr at 120 ft, after 180 days decay time. A handling cask providing a radial shield of 3 in. of lead will provide adequate personnel protection for the removal operation, if 180 days decay time is allowed before the trap is removed. An additional 2.4 in. of lead is required for offsite shipment of the cask. This additional shielding can be added after the trap is removed from the reactor building. Dose rates from the cold trap after the shield plug is removed from the access hole are shown. If direct line-ofsight exposure is avoided, dose rates to personnel will be below 100 mr/hr at any position, and below 10 mr/hr at distances greater than 20 ft from the access hole. Dose rates from the cask during its travel away from the hole, will be below 100 mr/hr at distances from the cask greater than 10 ft and below 10 mr/hr at 35 ft, if the cask is raised no more than 3 in. from the floor during its travel. Remote, unshielded handling of a primary sodium pump is feasible, since dose rates would be 100 mr/hr at 28 ft and 10 mr/hr at 90 ft, after ten years of operation, and providing that 14 days decay time is allowed to eliminate activity from the Na/sup 24/ film clinging to the pump. Dose rates after only one year of operation would be lower by a ...
Date: December 15, 1959
Creator: Rhoades, W. A.
Partner: UNT Libraries Government Documents Department