Space has been reserved near the southeast corner of the 100-N Area for the 122-N Decontamination Facility. Previous correspondence between Burns and Roe, Inc and General Electric bae discussed various facilities which might be needed in the building. The concepts of the decontamination processes are under active development by research groups at Hanford. At present, there are several workable processes known; each one has one or more fairly serious drawbacks.

In many cases of analysis of two-phase flow in systems, considerable computation or program time could be saved if the critical pressures ratio were known. If a reservoir or plenum pressure is fixed, the usual computational procedure involves the assumption of several critical pressures and the generation of several momentum terms to find the applicable critical pressure ratio and thereby the critical flow. The formulation of an equation of state make it possible to compute critical pressure ratios directly.

The amount of dry air or inert gas supplied to a "one pass" glove box ventilation system is proportional to the glove box inleakage. Most glove boxes in the 234r5 Building are equipped with sundry attachments, each contributing to the inleakage. No individual leak rate date is available for these auxiliary components in the "as installed" condition. Nor is the effect of time upon the leak rates known. Knowledge of these values, or at least an indication of the order of magnitude of the leakage attributable to each item, would provide a basis for analyzing glove box ventilation problems and for establishing criteria for new glove box designs.

A coolant line rupture during operation of a high temperature gas cooled graphite moderated reactor would present a serious hazard. The reactor would immediately depressurize and a great deal of air would be introduced into the coolant stream. As the air passed over the graphite moderator a runaway oxidation reaction would probably ensue unless an adequate safety system were available. This investigation was designed to evaluate chlorine as a reactor safeguard to be used to control a runaway reaction. Throughout this study, a small amount of chlorine in an air stream has demonstrated the ability to substantially reduce the oxidation rate of graphite. This has been the case even where the principal oxidizing agent was molecular oxygen or ozone.

Limitations on the service life of a fuel element imposed by degradation of the fissile core during irradiation have been a matter of great concern. Limitations imposed by changes in cladding properties during irradiation should be evaluated with equal care. Zircaloy-2, stainless steel, and aluminum alloys have been irradiated in the form of cladding on metallic and ceramic fuel elements. Several aspects of fuel behavior as influenced by these clad materials will be discussed. All observations related to irradiation behavior in this paper have been made on fuel specimens irradiated in water coolant.

Exponential and critical approach type measurements have been made to determine the critical mass, material buckling, and in a few cases, the extrapolation length for the lattices. This report attempts to list all measurements on water-uranium heterogenous lattices made to date at Hanford. All lattices were water moderated hexagonal arrays loaded with uranium of enrichments up to 3.15.

The reason for initiating the compact drive study for the T7 tanker was to investigate control rod drive size, location, and removal space requirement factors and select the control rod drive mechanism which would allow optimization of the over-all size of the containment vessel. Approximately twelve mechanical/hydraulic control rod drive arrangements were considered during this study.

A stimulation study of the deign as of June 1960 of the gas-cooled ORR Loop No. 1 was made using the ORNL analog computer. The proposed method of temperature control is evaluated, and the dynamic behavior of the loop for accidents and component failures is presented in graphical form.

Abstract: A definitive investigation of the boron-carbon equilibrium system is being made by X-ray diffraction, metallographic, and thermal analytical techniques. Additional baron-carbon alloys have been prepared by sintering and arc-melting compacts prepared from boron and high-purity graphite. Metallographic examinations of these alloys are in agreement with alloys previously preapred from lampblack. X-ray investigation of sintered compacts indicates that the solubility range of boron carbide extends almost to pure boron. Boron of various purities has been annealed for times up to four hours, but no structure other than beta-rhombohedral has been detected. Very high purity boron (10 ppm impurity) has been obtained for the study of allotropy and the equilibrium relationships at very dilute carbon contents.

Abstract: "As part of a program entitled "Improved Zirconium Alloys" a brief summary of a present zirconium research at various organizations has been prepared. Information was obtained by telephone and personnel contact; and these discussions, along with the organization and personnel, are presented."

An experiment has been performed using nuclear emulsions to measure the neutron flux and energy spectrum from a thick uranium target bombarded by 18-Mev electrons. The total integrated flux of 3.6 x 10(-4) neutrons/electron found here is considerably lower than the figure reported by other investigators. We attribute this discrepancy primarily to a wandering of the electron beam. The neutron spectrum should be unaffected. In addition to the expected peak at 1 Mev the energy spectrum shows a secondary peak at about 5 Mev. The shape of the experimental spectrum excluding this secondary peak, is fitted theoretically by a combination of the "evaporative" and "resonance direct" effects.

A modification of the Bridgeman anvils is described. This modification permits studies to be made to a load of 400 kbars, about twice that previously available.

The experimental demonstration of fission product release from over heated reactor fuel is necessarily subject in many respects to the arbitrary conditions imposed by the experimenter. Since an almost infinite latitude exists in the choice of materials, atmospheres, gas, velocities, temperatures etc., some allowance for an extrapolation to alternate conditions is definitely in order. The conditions imposed in this study are best described as those most likely to maximize fission product release. Two of the most important variables not investigated in the present report are the influence of metal cladding and the difference in internal nuclear heating as opposed to external radiant heating. In addition a significant uncertainty exists in the understanding of diffusion through large masses such as might result from a scaled-up melt down in a reactor.

Design analysis of the SM-2 core and vessel from work performed during 21 months.

SM-2 core and vessel design analysis.

SM-2 core and vessel design analysis.

"Graft polymers were made by Co60 mutual irradiation of mixtures of styrene and methyl acrylate in contact Teflon. These graft copolymers were analyzed by infra-red absorption and their composition determined. The composition of the graft copolymer formed from styrene and methyl acrylate was different from that which is predicted by the copolymer composition equation."

"The previously found maximum in the cosmic radiation energy spectrum is discussed, and the possibility that it arises in part from a misinterpretation of the data due to oversimplifying assumptions about the composition of the beam is examined. The importance that lack of recognition of isotopic composition may have is pointed out, and it is clear that this importance is strongly dependent on the measurements made and on the cut-off rigidity."

The objectives of this study have been to examine the problems of the pebble-bed reactor concept and to conceive a design of a facility for investigating the feasibility of this type of reactor. The design must provide for adequate leaktightness of the contaminated-gas system and adequate maintenance of contaminated components, the most vital feasibility questions of the concept.

Report regarding sixteen experiments performed with a miniature re-entrant gas loop. Includes design problems and difficulties encountered, as well as their solutions.

Abstract: This technical report describes the as-constructed SM-1 penetrated gasket designed for SM-1 Core and Flow Instrumentation (Task XIV). This report supplements APAE No. 79, The Summary Hazards Report for Task XIV, and evaluates the effects of a postulated failure of this gasket. The effects of failure on the Maximum Credible Accident are determined and conclusions and recommendations for the use of this gasket are made.

Report is a text revision of the International Temperature Scale of 1948; the numerical values of temperatures remain the same. The adjective "Practical" was added to the name by the International Committee on Weights and Measures. The scale continues to be based upon six fixed and reproducible equilibrium temperatures to which values have been assigned, and upon the same interpolation formulas relating temperatures to the indications of specified measuring instruments. Some changes have been made in the text to make the scale more reproducible than its predecessor. The triple point of water, with the value 0.01 ºC replaces the former ice point as a defining fixed point of the scale. It is also recommended that the zinc point, with the value 419.505 ºC, be used instead of the sulfur point. The recommendations include new information that has become available since 1948.

Abstract: Thermal characteristics of the SM-2 core were analyzed at steady state and loss of flow conditions. For steady state operation, the steady state code STDY-3 was used. For transients during a loss of flow accident, ART-02, a one-dimensional code, was used. This analysis indicates the SM-2 core is safe from burnout under steady state operation at design power level (28 tMW) because (1) no nucleate boiling exists, and (2) the minimum burnout ratio is above 2.0. The core is safe from burnout under loss of flow transient because the minimum burnout ratio in the hottest element channel of 1. 82 is above the minimum design criteria of 1. 5.

Report documenting the suitability of Louisiana for plant locations to produce synthetic liquid fuels, based on raw materials, water sources, and local interest.