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Activities in a program to determine the effects of neutron irradiation on A-302B steel are reported. Plans are discussed concerning performance tests on tensile, notched tensile, tube, and Charpy specimens. An irradiation capsule design was finalized. A theoretical analysis of the critical conditions arising in a tube subjected to both internal pressure and axial load is included. (J.R.D.)

Data are given on the radioactive nuclides present in waste disposal effluents during Nov. 1961. The concentrations of all activities was well within established limits for discharge to the environment. (C.H.)

This report provides a list of resources published in 1968 or 1969 related to poverty in the United States.

A survey of the pertinent literature was made to ascertain the status of data on U/sup 233/ fission-product yields. The various experimental determinations were evaluated, and the most recent mass-spectrometric results were used as a basis for deriving a set of preferred yields. These yields were compared with values reported in two other recent compilations, and for yields >1%, the three setrs agreed with each other to an average precision of <5%. It was concluded that recent measurements have somewhar improved the reliability of U/sup 233/ fission yield data, but some recommendations for additional experimental work were made (auth)

A design and operational description is presented of a BF/sub 3/ withdrawal mechanism planned for installation on one of the two startup channels at the SM-1. An analysis of possible malfunctions is included. (J.R.D.)

The sorption of the ions of uranium, copper, and nickel on hydrous zirconium oxide was investigated at temperatures from 25 to 250 deg C. The experiments were performed by equilibrating 5 ml of the test solution with 0.5 g of zirconium oxide in a titanium autoclave, which was heated by means of a rocking furnace. The sorption of uranium was affected by characteristics of the zirconium oxide, temperatare of equilibration, and concentrations of uranium and of free acid in the uranyl sulfate solutions. Conclusions are drawn concerning the relationship between each of these factors and uranium sorption. (auth)

Equations were derived by strain-energy methods for analyzing the stress and deflection in the boiler fuel element boxes used in the Pathfinder nuclear reactor. These boxes are of two types, both constructed of Zircaloy lI. One box is approximately 5 in. sq. The second is approximately 10 1/2 in. sq divided internally into four compartments (5 in. sq) by a cruciform control-rod guide structure. Stress and deflection in the box walls are caused by pressure differences across the walls. The equations were verified by testing an aluminum model of the single-element box and a Plexiglas model of the four-element box. There was less than 10% difference between theoretical and measured values for the maximum stress and deflection, and therefore, use of the equations for designing the Pathfinder fuel element boxes is justified. (auth)

A titanium pump loop, designed to circulate aqueous solutions at temperatures and pressures up to 370 deg C and 3000 psia, was constructed. It is to be used to study the chemical stability of uranyl sulfate fuel solutions of interest to the Fluid Fuels Reactor Program. The total loop voluime was minimized so that about 2 liters uf solution was sufficient for loop operation. The equipment includes a sampling system to remove solution samples from the loop while operating at elevated temperature and pressure; a hydroclone to separate and remove any solids and/or heavyphase material formed; and provisions for installation of corrosion test specimens in the main loop stream. All equipment performed satisfactorily at design conditions in tests with water. (auth)

The mathematical formulation of the subcritical reactor with extraneous sources is presented in detail. The solutions are presented for a two-group approximation including epithermal fission and the relationship to the usual critical reactor problem is discussed. Application of the equations to an arbitrarily complicated geometry is outlined, and an IBM-7090 program for the solution in a bare cylindrical reactor is described. (auth)

A criticality excursion occurred at the Oak Ridge Critical Experiments Laboratory on November l0, l961 as enriched uranium metal, neutron reflected and moderated by hydrogen, was being assembled. It is estimated that the energy yield was between 10/sup 15/ and 10/sup 16/ fissions. There was no personnel exposure or property damage. Fission product contamination, both airborne and contained in the metal, decayed sufficiently overnight to allow unhindered continuation of the experiment. The excursion was caused by a too rapid approach of the two sections of uranium constituting the experiment. (auth)

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Progress is reported on research and development work in pot calcination and radiant-heat spray calcination studies of synthetic Purex high-level wastes; and sorption studies using synthetic minerals and resins as well as natural minerals. The calcination studies are discussed in terms of batch calcination, melting of pot calcination products, spray calcination, and off-gas treatment; and sorption studies in terms of mineral reactions, fixation chemistry, and condensate wastes. (B.O.G.)

Two methods, the transpiration and direct P/sup 32/ counting methods, were tested for use in measuring the vapor pressure of TBP. With the transpiration technique, measured TBP vapor pressures were sufficiently reproducible, and probably of sufficient accuracy, to warrant improvement of the apparatus and its use in measuring the vapor pressure of TBP over various TBP- diluent-HNO/sub 3/-H//sub 2/O-UO/sub 2/ (NO/sub 3/)sub 2 solutions. With this method the vapor pressure of TBP containing 0.2 wt% H/sub 2/O was determined to be ~0.8 mu while that of TBP saturated with H/sub 2/O, ~6.5 wt% H/sub 2/O, was ~ 0.5l mu . The direct P/sup 32/counting technique was abandoned because of experimental difficulties. (auth)

A compilation is presented, in the twenty-five group Reactor Nuclear Data Tape format, of neutron cross sections for elements of major interest for GE- ANPD reactor analysis. The tabulated data are a reproduction of neutron cross section information contained on the Reactor Nuclear Data Tape, which was recently prepared. A brief outline of methods used in processing of the cross sections is also included. (auth)

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This report is a chronological Bibliography on Securities and Exchange Commission from 1962 through 1968.

The following report is part of a series of quarterly reports discussing investigations on the development of zirconium-base alloys having corrosion resistance and strength superior to Zircaloy-2 and/or development of materials of equivalent corrosion resistance by exhibiting enhanced strength. This report covers the period between May 29 and June 18 of 1958, made by the United States and the European Atomic Energy Community (EURATOM).

The discussion and summary report presented here illustrate the type of information that is currently in data processing machine language and available to those who require it. During the progress meetings at which these reports were discussed, several persons indicated a desire to have a documented explanation of the report and a personal copy to scan. This is the answer to these requests. The reports are comprised of three major sections. The first is a map indicating the number of recordings of data in the category defined by the function number code. The second is a summary of the data, in terms of hours, for each of the functions performed during an outage, except secondary function code number 21 which is the number of tubes associated with the basic function performed. The third is a summary of the secondary functions performed within a basic function class, the associated delay total, and the number of tubes involved for those functions where a delineation by tube count may be of interest.

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The discussion and summary report presented here illustrate the type of information that is currently in data processing machine language and available to those who require it. During the progress meetings at which these reports were discussed, several persons indicated a desire to have a documented explanation of the report and a personal copy to scan. This is the answer to these requests. The reports are comprised of three major sections. The first is a map indicating the number of recordings of data in the category defined by the function number code. The second is a summary of the data, in terms of hours, for each of the functions performed during an outage, except secondary function code number 21 which is the number of tubes associated with the basic function performed. The third is a summary of the secondary functions performed within a basic function class, the associated delay total, and the number of tubes involved for those functions where a delineation by tube count may be of interest.

This document details activities of the Irradiation Processing Department during the month of October 1964.

About 298,000 OIIIN aluminum components were received from Harvey for evaluation after major steps were take in the vendor`s plant to upgrade product quality. This material met HAPO specifications and was comparable to Alcoa control components. Finished fuel elements (non-bumper and bumper) canned in the test and control components are tabulated by uranium core type and canning month. Non-destructive and destructive test data did not show any differences in clad fuel elements attributable to component quality. Irradiation status of non-bumper and bumper fuel elements clad in Harvey improved components as of Dec. 31, 1962, is tabulated, as is a summary of ruptures. It was recommended that Harvey Aluminum Co. be allowed to bit competitively on all future Al component orders except for the KVN and KVE component models.

The objectives of this Process Improvement Transition Authorization (PITA) for KE and KW reactors are to authorize a flat 700 C graphite temperature limit in the bulk of the graphite, with a conditional 850 C limit in the 0.130 coring zone only, and in conjunction therewith to provide for routine use of nitrogen instead of carbon dioxide in the reactor atmosphere. These temperatures as quoted apply to filler blocks in the vicinity of normal channels, and are maximum limits.