The need for special high velocity shock testing is pointed out and equipment that has been developed to meet these needs is discussed. The concept of velocity-change and its importance in shock-testing technology is brought out. The three types of equipment discussed are: accelerated drop testers, pneumatic actuators, and air guns. Examples of each type and their capabilities and limitations are presented.

From foreword: The purpose of this conceptual design is to outline criteria for the selection and development of a reactor plant with a fossil fuel fired superheater, a conventional generator plant and auxilary systems constituting a complete operating unit capable of producing a gross electrical power output of 22,000 kw.

Abstract: A discussion of the potential hazards of chlorine trifluoride is presented and a series of safe handling procedures is recommended.

Tory 11-A, the first experimental reactor in the Pluto nuclear ramjet program, will be tested in late 1960 at the Nevada Test Site of the Atomic Energy Commission. The fundamental objective of Tory II-A is to demonstrate that a high power density, high-temperature, air-cooled reactor can be successfully designed, constructed, and operated. This application places requirements on the reactor control system which are considerably more stringent than those found in previous reactor systems, both from the standpoint of radiation environment and system performance. To fulfill the system requirements a high-performance reactor control system has been designed and built; control actuation hardware has been developed which can withstand the high-radiation environment of Tory LI-A. This actuation system features radiation-tolerant electrohydraulic components: actuators, servo-valves, solenoid valves, feedback transducers, accumulators and associated hydraulic components. To provide high reliability against undesired shutdowns while carrying out high power level transients, the Tory 11-A control philosophy includes a new concept in reactor safety: reliance on a nonlocking, fast-reset safety system during accident situations. The fast-reset system acts so as to hold the programmed power level rather than initiating an irreversible "scram" action. A reactor "scram", if tolerated, could produce thermal shocks which would seriously damage the …

This report is a corrected copy of the authors "rapporteur talk" at the 1959 Kiev Conference on High Energy Physics. It contains data on the rapporteur system and on interactions of strange particles with protons and neutrons, using data from bubble chambers, counters, and emulsions.

A program has begun embodying a study of the diffusion rates of krypton and xenon into unirradiated UO2 and UC at temperatures between 1000°C and 2500°C.The experimental technique is described in some detail, involving exposure of UO2 and UC samples to a Kr85 atmosphere at various temperatures, removal of surface-adsorbed krypton, dissolution of the sample, collection and ion chamber measurement of the diffused Kr, and calculation of diffusion coefficients from these data.

Report documenting ongoing research and developments at the Oak Ridge National Laboratory's Gas-Cooled Reactor Project.

Report describing the Hanford Laboratory controlled potential coulometer. This includes the theory behind the coulometer, details of its circuitry, and performance tests. Appendix begins on page 23.

This paper for the Project Vela Technical Symposium is intended to be a brief resume of the principle results of the Cowboy program. The material is taken from a report entitled "Use of Large Cavities to Reduce Seismic Waves from Underground Explosions" by Herbat, Werth, and Spring (1960). As a result, no effect is made to justify each step in the analysis; reference should be made to the complete report.

Abstract: Development of welding processes for stainless steel fuel elements, previously initiated by Alco Products, was continued in Task 5. Three objectives were accomplished; (1) development of a suitable reference welding process, (2) evaluation by out-of-pile tests of the structural integrity of reference welded fuel elements, and (3) preparation of specifications and fabrication procedures for the reference welding process.

The components for the present aluminum clad, Al-Si bonded, internally and externally cooled (I & E), uranium fuel elements are composed of impact extruded cans and spire caps as shown in Figure 1. This type of component requires two impact extrusions; however, in December, 1957, J. E. Ruffin proposed another design of component in which there was only one impact extrusion. For this component, Figure 2, the spire was impact extruded as a part of the can.

The large scale application of Zircaloy-2 pressure tubes for structural use either in or out of reactor service, is without precedent. For more common materials, there normally are adequate data and long operating histories on which to base design and service limits. In the absence of such information for Zr-2, several investigative programs have been devised to provide much of the information from which design and service limits may be defined for Zr-2 pressure tubes. These investigations encompass in-and-out-of-reactor creep and stress-rupture testing, pre-and-post irradiation testing, and bust strength, as well as the effect of flaws or defects (from both fabrication and service origins) on burst strength and fracture characteristics. Already creep and stress rupture testing of unirradiated Zircaloy-2 is well advanced, and some experimental pre-irradiation burst testing has been carried out and will be extended rapidly as improved equipment becomes available. One irradiated KER tube sample has been burst tested and the requirement for post irradiation burst testing equipment have been defined.

Over the last decade, several computational methods have been developed and used to examine reactor flow transients caused by pump outage. The variations in system character which have been analyzed are sufficiently diverse that it appears worthwhile to compile them into a single report.

A portable shield has been designed, developed, fabricated and shop tested to provide the HRT with a facility for direct dry maintenance operations. It provides temporary replacement for any one of the lower roof plugs and should permit many operations to be performed without flooding the reactor cell with water.

The effect of adding small quantities of fuel or poison to the HRT has been estimated using perturbation theory. The results have been reduced to a single relation and a set of graphs which make the estimation of added reactivity relatively simple. The perturbation theory results are compared with multigroup results and reasonable agreement is demonstrated; however, there is some question concerning the prompt neutron lifetime.