EWR-ESE-144, Fifty MC-890 type final assemblies were subjected to irradiation at low temperature test in order to determine if irradiation would reduce the frequency of high voltage breakdowns. A control lot of a like number of units was subjected to low temperature test in the same manner except without irradiation.

Methods for the noncorrosive removal of volatile fission products and UO2 dust from carbon steel and stainless steel have been developed. Procedures for applying these methods to the decontamination of the EGCR charge and service machines are described.

Considerable work has been done on decontamination procedures for both corrosion products and fission products. Testing of promising procedures under typical rupture conditions is an important phase of this work. The Irradiated Rupture Prototype (IRP) has been used since August in this evaluation work. This document is one of a series reporting these data.

A conductometric titration method described by Goldstein was adapted for use in a remote analytical facility. The results obtained by mean of experiments made prior to this adaptation indicated that methanol is the most satisfactory medium in which to determine excess sulfuric acid in uranyl sulfate solutions that stimulate Homogeneous Reactor type fuel. When methanol is used, the complexation of hydrolyzable ions with sodium fluoride, as described by Pepkowitz, Sabol, and Dustin, is not required.

Answers by ICPP personnel to questions asked by Eurochemic are given on a variety of subjects, including off-gas sampling systems, off-gas filtration, iodine evolution during fuel dissolution, and process cell contamination. Data are presented on the% of I/sup 131/ retained in waste solutions in the absence and presence of mercury. (D.L.C.)

An experimental investigation into the gas dynamics of a jet-driven vortex tube for application of a cavity nuclear reactor to rocket propulsion has shown that viscous retardation of the vortex motion is severe, because of a high level of turbulence near the periphery. Based on the experience gained in this study, it is estimated that the achievement of vortex strengths sufficient for practical application will require the use of small diameter tubes with appreciable expenditure of power for recirculation of the gas. The effect of the high degree of turbulence on the separation process near the periphery remains to be determined. The independent variables which were found to influence the vortex strength significantly for a given gas and temperature condition are the tube diameter, the mass flow rate per unit tube length, the injection velocity, and the wall pressure. Estimates of the degree of turbulence in vortex flow have been made from data on the variation in tangential velocity with radius. Virtual (total) viscosities near the periphery ranged from 30 to 700 times the molecular viscosity for tangential Reynolds numbers of from 4 x 10/sup 4/ to 2 x l0/sup 6/. Measurements of the position of the mole- fraction peak …

EXPIRE is a calculation which predicts the reactivity-lifetime, instantaneous and integrated effective multiplication constants and instantaneous and integrated effective multiplication constants and instantaneous conversion ratio for heterogeneous reactors. The concentration of all the isotopes of interest from Th232 to Am243 are calculated as a function of time using the average reactor power density and a uniform flux distribution. The equations have been programmed for the IBM-704 computer and the average running time is approximately two minutes per reactor lifetime.

Release of fission products from UO2 with emphasis on fuel elements operated at higher surface temperatures and lower external pressures than those for pressurized-water systems.

This monthly document details activities of the Hanford Operations Office during the month of March 1960. (FI)

This monthly document details activities of the Hanford Operations Office during the month of May 1960. (FI)

Abstract: The purpose of this report is to describe the problems inherent to general dielectric strength measurements, and to act as a guide in establishing a method for pulse dielectric strength measurements.

This report discusses the immigration policy of the united states before the immigration and national act of 1952.

Multigroup one-dimensional calculations were done recently to obtain estimates of critical masses, power density distributions and fissioning spectra for some homogeneous molten salt reactors having outer reflectors and central "islands," placed inside the currently proposed MSRE vessel. For a 5-inch-thick outer reflector and 1-ft-diamter island, both beryllium, the calculated critical mass is 108 kg; 40 percent of the fissions occur at thermal, and the maximum power density of 3.9 times the core mean power density occurs at the island-salt interface. If the reflector thickness is increased to 10 inches, the critical mass is reduced to 34 kg; 67 percent of the fissions occur at thermal, and the peak power density of twice the core mean again occurs at the core island-salt interface.

This report discusses the isothermal hydraulic demand characteristics were determined by laboratory experiment for full charges of self-supported I&E fuel elements in a zirconium process tube. Pressure drop, flow rate data, and the calculations of annulus-to-hole flow ratio are presented. For self-supported fuel elements, pressure drop does not vary with temperature as much as it dies for non-self-supported furl elements.

9 > ? ( ing schedule included MTR, ETR, and other aluminum-type fuels that were processed through the TBP headend at 75 to 100% of flowsheet rates and through the Hexone second and third cycle extraction system at up to l5O% of flowsheet values, achieving an over-all recovery of 99.60%. In carrying out basic studies on aqueous zirconium processing, results obtained during the potentiometric titration of zirconium-fluoride-nitric acid solutions with sodium hydroxide were explained by a mechanism in which the untitrated zirconium fluoride species is converted to a fluozirconate and further hydrolysis proceeds via a fluozirconate route. A modified Zirflex flowsheet for processing 2.5% U-Zr alloy fuels is proposed as a result of bench scale dissolution studies on unirradiated PWR plates that showed that dissolver solutions could be stabilized by adding aluminum nitrate and nitric acid. that higher ammonium fluoride concentrations increased the dissolution rate, that higher hydrogen peroxide concentrations lowered the dissolution rate and decreased solution stability, and that ammonia removal by boiling with or without air sparging had little effect on the reaction. Density curves for Zirflex solutions are reported. First tests of the pilot plant for continuous dissolution of zirconium-type elements have indicated that the dissolution rate …

Sugar Stock and Sales report for Imperial Sugar Company including a list of sugar sales for the date showing the amounts and estimated balance separated by location, raw sugar, and refined sugar sales.

The objective of this report is to summarize all in-reactor operating data associated with PT-288-A. This document will serve as a ready reference to the irradiation history of the fuel elements, and will record any anomalies encountered in the performance of this test.

The loop was charged October 30, 1959 with seven 12-inch Natural Uranium Zircaloy-2 clad 7-rod clusters. The test was primarily for the new hot-headed method of end closure used on these elements. The loop was pressure tested at 4000 psi after it was charged. Initially the loop was held at low-temperature to study the buildup of oxygen (presumably from radiolytic decomposition of water). After startup the neutron activity held at slightly above normal but the strainer gamma activity was exceptionally low. Frequent additions of LiOH bombs were necessary to maintain the pH at 10.0 (previously it was 4--5 pH but was raised to pH 10 for this test.) After the temperature was raised to operating conditions the pH held nicely at 10. On November 16, 1959, the heat exchanger exit temperature thermocouple blew out resulting in depressurization of the loop and a reactor scram. Repairs were made during the outage and the loop was returned to normal operation. Eleven scrams occurred during the test period caused by 105-KE or 1706-KER.

A procedure is reported for the sea disposal. by fixation in concrete. of ion-exchange resins containing sorbed fission products and filtered corrosion products. The results of sea disposal of both treated and untreated resins is presented. For the fixation of nuclear-reactor-system demineralizer resin in cement, a mixture of 2 parts of cement, 1 part of resin, and 1 part of water was found to be optimum. A concrete block prepared by this receipt showed only 0.44% of the gross gamma activity leached after 2540 hr of contact with simulated sea water. The data indicate that very little or no further leaching will occur. (auth)

None

Research efforts in neutron physics are reported in the areas of light- water-moderated reactors, new reactors for BSF and TSF, critical experiments, experimental nuclear physics, experimental reactor physics, reactor theory, shielding research, radiation detector studies, plasma physics theory, and theoretical nuclear physics. Separate abstracts were prepared for the seventy sections of the report. (W.D.M.)

A discussion of the hazards and philosophy of plutonium handling is presented. Glove box construction and materials are also discussed along with handling techniques and work being done with plutonium in various parts of this country. (J.R.D.)

The objective of the production test detailed in this report is to evaluate the behavior of N-Reactor inner fuel tubes during irradiation in the KER Loops at conditions equivalent to those expected in the N-Reactor. Natural and .947% enriched Zircaloy-2 jacketed N-Reactor inner fuel tubes (NIN-1) and (NIE-1) will be irradiated in the KER Loops to exposures no greater than 3500 MWD/T. The elements have brazed end closures and Zircaloy-2 supports and are contained in Zr-2 flow distributing sleeves.

Abstract: This technical report is concerned with the program and results of the SM-1 reactor vessel head removal and core inspection at Fort Belvoir, Virginia from the period March 7, 1969 to May 17, 1959. It covers the operating procedures in detail and records the conditions found and problems encountered in order to make a record for reference in future work of this nature. The major objective of the program, to obtain irradiation data on the SM-1 type core, has been met. The boron control rod elements were found unsatisfactory for full core life, and europium oxide elements were placed in the core for future irradiation stability data. A major problem was experienced with the cracked pressure vessel head studs. The methods developed for the removal of the broken studs are presented. The complete metallurgical study of the stud failure from stress corrosion is included as Appendix B.