AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 10, 1952 (ORNL-1227)
R. C. Briant, Director
A. J. Miller, Assistant Director
Edited by W. B. Cottrell
MARCH 10, 1952
This is the quarterly progress report of the Aircraft Nuclear Propulsion Proj~ct at the Oak Ridge National Laboratory and summarizes the technical progress· on the project du ring the period covered. It inc lud.es not only the work of the Laboratory under its own contract, W~7405-eng-26, but also the research for the national ANP program performed by Laboratory personnel. The report is divided into four parts: I. Reactor Theory and Design; II. Shielding Research; III. Materials Research; and IV. Appendixes. Each part may be regarded as a separate entity and has a separate “Summary and Introduction.
Analysis of the circulating-fuel aircraft reactor has been extended to systems incorporating intermediate heat exchangers, various secondary coolants, liquid moderators, and the use of heavier reactor shielding (sec. 1). All these systems utilize the fundamental advantage of the bi- functional fuel-coolant, and appear to be capable of supersonic nuclear propulsion. The location of the heat exchangers around the reactor results in lower shield weight, even with a larger shielded-volume diameter, than a tandem reactor and heat exchanger arrangement. In order to perform a limited amount of aircraft maintenance without special shielding, various modifications of the minimum divided shield specifications have been investigated.
Studies of the performance and design of the circulating-fuel air- craft reactor are sufficiently en- couraging that the first Aircraft Reactor Experiment (ARE) to be constructed by the Oak Ridge National Laboratory will be of this type (sec. 2). The reactor core, as designed for the ARE, consists of a beryllium oxide moderator with a multipass fuel-coolant system. The core and a surrounding beryllium oxide reflector are contained in an Inconel pressure shell. Design of the reactor, fluid circuits, building, and associated equipment are essentially complete. The reactor is expected to he in operation early in 1953.
3). The techniques of the preparation, purification, and handling of the fluoride mixtures have been developed so that 100-lb batches of the treated fluoride may be prepared and loaded in adequately cleaned test equipment. Techniques of pumping, sealing, and controlling the fluoride coolants and lubricating moving parts of the systems have been demonstrated I at temperatures above 1300°F, and it is considered that these techniques are adequate for ARE application.
A centrifugal-flow fluoride pump has I operated for weeks with neither mechanical failure nor leakage. Liquid sodium technology appears to be well in hand, since continued success has been experienced in the operation of sodium (or NaK) pumps, seals, and heat exchangers. The NaK-to-NaK heat exchanger loop has now operated for 2300 hr with a maximum temperature of 1500°F. Gross heat transfer studies indicate that space-economical systems and components can be built to handle copious quantities of heat, as required by fluoride systems, at temperatures between 1200 and 1800°F.
The reactor physics calculations, which have further defined the statics of the circulating-fuel ARE, have led to some general observations regarding the kinetics of both the circulating- fuel ARE and ANP reactor (sec. 4). Although the thrombosis effect is an important concern in the control of these reactors, the loss of the delayed neutrons may not be if the circulation of the fuel itself is as good a damping mechanism as now indicated. These The developmental work in reactor kinetic difficulties are of less con- plumbing and associated hardware has cern to the ARE than to the ANP, since been primarily concerned with the the circulation rate in the ARE is so technology of high-temperature fluoride slow that the control rods can cope mixtures, and a secondary effort has with the thrombosis effect and a large been the study of liquid metals (sec. fraction of the delayed neutrons are emitted into the active volume. The current ARE design has a critical mass of 22.3 lb, a total uranium investment of 74 lb, 71% thermal fissions, and a leakage-to-absorption ratio of about 1 to 3. Brief studies of hydroxide moderated reactors (including KOH, LiOH, NaOH, RbOH, and SrOH) show that, KOH, the hydroxide moderated reactors require low critical masses and small core volumes for minimum critical mass.
Measurements on the critical experiment of the simulated General Electric direct-cycle reactor have been completed and the simulated circulating-fuel reactor is now being assembled (sec. 5). Evaluations, in terms of contributions to reactivity, have been made of several reflector , modifications of the direct-cycle assembly. In addition, the data from except for the earlier graphite reactor assembly have been correlated with the data from theoretical calculations of the assembly. The correlation lacks precision but gives results that are at least consistent with the experimental facts.