Equilibrium krypton and xenon adsorption isotherms were obtained for four varieties of charcoal and for Linde Molecular Sieves Types 4A and 5A, generally at 0, 25, and 60°c. Such data are of interest in connection with design and evaluation of adsorbers for radioactive noble gas fission products. The isotherms were fitted, by linear regression analysis, to straight-line forms of the Freundlich and Langmuir equations. The Freundlich linear equation gave the better fit and the parameters of this equation are presented for each of the isotherms. Also presented are the constants for an equation representing the temperature dependence of arbitrary adsorption coefficients, the coefficients having been calculated from the Freundlich isotherm parameters. Some aspects of the applicability and accuracy of these results are discussed.

The general solution to the basic differential equation d^4w/dy^4 +4w= -4f(y) is transformed from the primary form treated in most texts to an alternate form in which each integration constant corresponds to one edge condition at y=0. The relationships between the integration constants of the two forms are derived and the values for the transformed functions are tabulated. The particular solution is derived in general and given in unique form for various functional forms of f(y). Matrix notation is used throughout the derivations; however, a knowledge of matrix theory is not need for application of the results.

One of the techniques by which highly ionized plasmas can be generated in the laboratory makes use of strong, electromagnetically driven shock waves propagating into a cold gas. In this paper the phenomenon is analyzed as a one-dimensional single-fluid hydromagnetic problem, neglecting dissipation behind the wave. We hypothesize that the rarefaction wave remains attached to the front. In the limit of essentially complete ionization behind the front the problem can be solved analytically as long as the transverse magnetic field there remains small compared with the longitudinal field.

One of the characteristics of the safety control system of the Hanford reactors which is of primary concern from the standpoint of reactor safety is the response time. The purpose of this study is to determine whether the safety control system is sufficiently fast-acting to prevent an intolerably large power excursion from developing. An intolerably large power excursion is one that would cause a melting of the uranium while the reactor power is above the previous operating level. The purpose of this report is to show the techniques used to analyze the speed of control of the Hanford reactors. This report will help to standardize the methods and reduce to a routine the solution of this type of problem. The analog circuits will be presented and explained. A report has been written previously on the speed of control. The changes in the problem have necessitated the redesigning of the analog circuit and also the issuing of this report. This report compiles and organizes the analog work done on the speed of control to date.

None

This report provides operational worksheets detailing charge/discharge, maintenance, fuel element ruptures, leaks, tube replacements, and production data.

This report consists of eight curves, showing Plutonium-240 concentration and conversion ratio as a function of graphite temperature for various goal exposures. A range of 450 to 1000 degrees centigrade on the moderator temperature is used. Other parameters are consistent with the methods and constants used in present product accountability. No attempt has been made to state absolute conversion ratios or Pu-240 percentages. The purpose of this report is to provide a basis for use in economic studies involving changes in graphite temperatures. No conclusions as to operating changes should be inferred from these curves since the combined economics of any change in graphite temperatures include significant and complex factors in addition to conversion ratio and product quality consideration.

Since start-up of the neptunium recovery and purification program at HAPO, the neptunium-has been subjected to the following process steps: Production of crude neptunium nitrate by the Redox and Purex plants; Production of purified neptunium nitrate, using an anion exchange process, in research and development facilities operated by CPD`s Process Chemistry Operation; Production of neptunium oxide (NpO{sub 2}) by Process Chemistry. Mixing the NpO{sub 2} powder with aluminum powder and incorporation in an aluminum-clad-target element, done by the Plutonium Metallurgy Operation of Hanford Laboratories; and Irradiation of the target elements in the Savannah River reactors; and Chemical reprocessing of the irradiated target elements at Savannah River to recover the desired plutonium-238 and the unconverted neptunium. As frequently happens to between-contractor transfers, neptunium material balances between Hanford`s NpO{sub 2} and Savannah River`s recovered neptunium have sometimes been poorer than expected. Because of wide spread interest in this problem, it was deemed desirable to put on record a description of the measurement techniques which have been used at Hanford for NpO{sub 2} and the improvements which have been made in the analytical methods. The measurement methods used for purified nitrate are also described.

This PITA is designed to permit large-scale application of a change in process water inhibitor concentration before a permanent change to the Water Plant -- Process Standards is fully justified and implemented. Specifically, this PITA authorizes reducing process water dichromate concentration to 1.0 ppm at all reactors (C excluded), provided pH remains at 7.0 {plus_minus} .1.

Figures are included which gives the necessary information on C Reactor for analysis of stress buildup in the graphite tube blocks.

The present Quarterly Report is the continuation of a series issued by the new Fuels Development Operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations.

This document details activities of the irradiation processing department during the month of March, 1961. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; Financial Operation; and NPR project.

This document details activities of the irradiation processing department during the month of July 1961. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; Financial Operation; and NPR Project.

This document summarizes the amounts of radioactive contamination discharged to ground from separations facilities through December 1960. Detailed data for individual disposal sites are presented on a month-to-month basis for the period of July through December 1960.

During the first two weeks of May while PRTR was being tested at a power level of a few MW, rather large amounts of some fission product gases were observed in the helium system. Samples of D{sub 2}O from the primary cooling system taken during the second week of May were analyzed and found to contain I{sup 131-133-135} and some fission gases (and their daughters) but practically no other fission products. On May 14 the reactor power level was raised to 10 MW and operated at about that level until May 18 at which time the reactor went down. (The power level was actually down to 2 MW for a short period late on May 17, then operated at 15 MW until 1620 on May 18.) On May 17, samples from the primary D{sub 2}O and helium systems were obtained and quantitative analyses of several of the short-lived fission gases and other radioisotopes were performed. Analyses of these samples showed a high concentration of fission gases in the helium system and high I{sup 131} in primary D{sub 2}O system and its ion exchange bed, suggesting possible leakage from a fuel element. Chemical analysis of the primary D{sub 2}O system showed …

Abstract: "This is the first interim report of a literature survey investigating the uses and technology of radioscopes in the countries of the world, with particular emphasis necessarily placed on those more highly industrialized nations. Part I includes a general summary of the quantity and quality of radioisotope work being performed in these countries. Among the outstanding radioisotope applications with respect to quantity is the work of Japan in chemical research, that of USSR in metals research, and that in the USA in medical research. A detailed qualitative examination of the relatively narrow region of radioisotope applications in leak detection is also presented. Part II of this report gives a direct comparison between the progress and status of the US and the USSR in the varied uses of radio-isotopes."

A tabulation is presented of values for selected properties of lithium hydride. (M.C.G.)

A remote dissolution and analytical program for irradiated thorium is given. The aluminum jacket on the slug was dissolved with 6M nitric acid and 0.005M mercuric nitrate. After a water wash, the thorium dissolution was accomplished with concentrated nitric acid made 0.04M in hydrofluoric acid. Weighing, dissolving, and sampling were done remotely in the multicurie cell at the Idaho Chemical Processing Plant. Handling techniques for weighing and dissolving the slugs are described. Transferring and sampling apparatus as well as sampling techniques for the dissolved material are discussed. Analytical data obtained are tabulated. Abstracts of analytical methods for uranium concentration and isotope ratio, aluminum, thorium, cesium, and cerium are given. (auth)

A philosophy for designing wideband multistage transistor amplifiers is presented. The amplifier was visualized as an integral unit, the interstage networks constituting the elements of the amplifier unit. By designing the amplifier as a unit and adjusting the over-all response (gain and bandwidth) with the interstage time constants, an increase in gain-bandwidth product was realized over the iteratively designed amplifiers. The resulting increase in gainbandwidth product resulted from absence of the bandwidth shrinkage factor for multistage amplifiers. Formulas were derived for both a two- and three- transistor integrally designed wideband amplifier, in which shunt peaking networks were used for coupling. Experimental amplifiers were con structcd following these formulas, and the observed performance agreed quite well with the calculations. (auth)

The estimated achieved reliability of SNAP 10A space nuclear power units will be relatively low at the timeof the first SNAPSHOT flight test in April 1963 and the existing R&D program does not provide a significant reliabiity growth thereafter. The total costs of an 8-satellite network using SNAP 10A units over a 5-year period has been approximated for the case where the total cost of a single satellite launched is 8 million dollars.

Tests were conducted to determine the direction and magnitude of any short- or long-term drift in core thermocouples, primary loop and pressurizer resistance thermometers, and resistance thermometers located in the secondary side of the boilers for Seed 2 at 2303 EFPH. It was found that drift in the core thermocouples is definite but random. Other findings are discussed and calibration data are presented graphically. (J.R.D.)

A test was run on the Shippingport Pressurized Water Reactor on June 15, 1960, to determine whether or not the pressure snd water flask volume of the valve operating system are sufficient to meet requirements under both normal and emergency conditions. The system was found to be satisfactory. A two-hour leak test indicated leakage. (D.L.C.)

A survey of the character and magnitude of hazards involved in processing liquid metal bonded fuels was made and the scope of a preliminary experimental program outlined. Processing of SRE and CPPD fuels by mechanical decladding followed by controlled reaction of the collected methods. Simdlarly, shearing of PRDC fuel and controlled exposure of the Na in the severed portions to water appears more desfrable than chethical dissolution of the metallic cladding. (auth)

An analysis of the EGCR core structure was made to determine the lateral deflections (bowing) of the graphite columns resulting from shrinkage caused by fast-neutron irradiation, the life expectancy of each column due to restraints imposed on the bowing, and the reaction forces induced in the supporting structures. Based on currently avallable data for EGCR type graphite shrinkage and assuming experimental loop operation, a maximum bowing potential of 3.61 in. was calculated for an interior column. It was found that strains equivalent to the rupture strains observed from tensile tests could be expected after 4 to 6 years of full-power operation. Over half of the columns will reach these strains before the 20-yr reactor design life is reached. (auth)