None

The substitution of Al-clad for Zr-clad fuel necessitates reducing the operating temperature. The amount of saving that must be made in fabrication and chemical reprocessing to overcome the cost of operating at reduced thermal efficiency is estimated for a boiling water reactor. (auth)

The power and estimated reprocessing load is tabulated for existing and proposed US and UK built non-military reactors over 10 kev. (auth)

The UCRL light-pulse generator provides flashes of light accompanied by electrical pulses generated at an impedance level of 50 ohms. The usual repetition rate is 60 cps. The 10 to 90% rise, both for the light and for the electrical pulse, is less than 0.5 m mu sec. The light pulse rises and then falls to 50% peak amplitude in less than 1.5 m mu sec. The time relation between the light and electrical pulses is fixed. The electrical pulse power available is large. In conjunction with an oscilloscope, the light-pulse generator has proved useful in the testing and evaluation of multiplier phototubes and lowlight- level image devices. Time resolution of 10/sup -10/ second is typical of measurements made with conventional fast oscilloscopes (e.g., Tektronix 517), while elaboration of technique permits relative time measurements that are better in some cases by at least three orders of magnitude. The light is emitted from a region a few mils in diameter, and thus may often be considered to come from a poirt source. An S4 photosurface subtending 0.1 steradian at the light source emits photoelectrically about 10/sup7/ electrons per steradian per pulse. For convenience, three decoupled electrical out-put channels are provided, together …

The radioactive fission products formed during irradiation of the uranium fuel in the reactors are separated from the uranium and plutonium in the Chemical Processing Department`s separations plants. Most of these fission products are removed in aqueous solutions as high-level radioactive waste and are subsequently stored in large underground tanks. The purpose of this report is to record the history of HAPO process waste storage, together with the developments which have significantly affected waste storage volumes and costs since start-up in 1944.

Oxide and carbide coatings and methods of applying them to graphite crucibles were evaluated on a laboratory scale. The objectives were to find a better crucible and mold coating than magnesia (Mgo), which is currently used in the remelt operation, and a better method of application than brushing. The evaluation consisted of the vacuum-melting of slug uranium charges and derby having different coatings, followed by studies of the appearance of the coatings and the metal quality (impurity levels) of the products. Flame-sprayed magnesium zirconate (MgZro/sub 3/) was determined to be the best coating evaluated. It limited the average carbon pickup to 73 ppm for derby uranium charges; with slug uranium, no pickup occurred. The flame-spray method provided a uniform coating thickness and virtually eliminated moisture and volatiles in the coatings. Studies showed that the flamesprayed magnesium zirconate contains a very small amount of volatiles and is not hygroscopic, but that 2500 deg F it is reduced slightly by graphite and co is released. (auth)

References to signiflcant design information for a steam generator in a gas-cooled reactor system are presented. Papers concerning heat transfer and pressure drop correlations for gas flow across banks of bare and circular finned tubes and for subcooled waters boiling water, and superheated steam flow inside pipes are listed and briefly discussed. (auth)

The design philosophy, control actions, and functional operation of the instrumentation and control associated with the ORNL-20 in-pile experiment operated in the MTR in 1955-6 are briefly described. (auth)

The several incidents of ozone explosions in cryogenic apparatus in radiation fields are reviewed. It is concluded that the explosions may be avoided if oxygen regions where, in static systems they tend to concentrate by liquefaction. Such provisions would make the proposed circulating N/sub 2/- cooled trap for the N.S. Savannah off-gas system acceptable if the initial oxygen content of the nitrogen coolant is less than 200 ppm. Higher concentrations are undoubtedly safe but the lack of information on such facets of the phenomena as quantities of ozones required radiation exposure, and explosion- inciting agents does not permit a quantitative definition of the limiting conditions for an explosion. (auth)

A series of im-pile fuel sample irradiation experiments was initiated to obtaim radiation damage information on fuel material combinations applicable to Foreign Reactor Program Fuel Elements where U/sup 235/ enrichment is limited to 20%. The series of expcriments will be divided into several phases, each being based on the materials combination employed. This report describes Phase I of the program which consists of determining the integrity and suitability of a 48 wt.% U--3 wt.% Si-49 wt.% Al alloy as a fuel for Foreign Reactor Fuel Elements. (auth)

This document discusses the standards and testing relating to the allowable water shutoff times for reactor tubes and crossheaders are based on theoretical calculations. The resulting limits are believed to be quite conservative and to restrict operations unduly in many instances. They are restrictive because of the conductivity and resistance data on which they are based. It is the purpose of the test to obtain the experimental data necessary to determine new water shut-off curves for the reactors. When reliable information is obtained for elevated temperature lattice conductance then theoretical calculations applying to those values will be far more accurate. The information will be related to water shut-off times, reactor safety considerations such as rod strength and drop time versus uranium fuel melting, etc. It can be roughly estimated that this information will reduce the time delay before tube and/or cross-header water shut-off by as much as 25%. This will save approximately one to two days production per year or about $60,000.

Four Zircaloy-2 jacketed enriched seven-rod cluster elements, one three-foot three-rod Zircaloy-2 jacketed enriched cluster element with integral end closures, and two eighteen-inch Zircaloy-2 jacketed wire wrapped cluster elements containing UO{sub 2} will be irradiated at jacket surface temperatures up to 270 C in the KER loops to an exposure of 3500 MWD/T on the enriched seven-rod cluster elements.

Iron oxide deposits removed from high pressure system of HRT after the first power run will be annlyzed for Fe/sup 59/ in an attempt to determine the time these deposits were circulated prior to deposition (settling or film formation). Deposits available are film from corrosion specimens (stainless steel, zirconium and titanium) and settled particles. (auth)

The performance, reliability, and reactivity lifetime of the Shippingport PWR are determined under normal steady state conditions. These characteristics are studied over a period of 1011 EFPH, at a power level of about 60 Mw(e). Flux tilts that occurred during this time, as well as the means employed for their suppression, are described. (T.F.H.)

Abstract: Scouting experiments indicated that calcination of highly active Darex and Sulfex decladding and Purex extraction wastes will not release hazardous amounts of fission products to the atmosphere. Simulated wastes containing up to 90 curies/liter or activity were evaporated and calcined to 750°C and the off-cases passed through a condenser and a series of caustic scrubbers. Less than 0.1% of the fission products was released, of which ruthenium was 10-90%. Fission product release was lower from neutralized than from acidic wastes. Fission products were leached appreciably, 0.05% of the beta and 0.02% of the gamma, from the calcination solid with water in 96 hr.

On occasion a rear pigtail of a reactor fails and is blown free from its connectors. The coolant water then discharges into the rear face area. This in itself is not a particularly hazardous condition and need not necessarily require a reactor shutdown. It would be desirable, then, to continue operating the reactor in this condition until a shutdown is convenient. A question which does arise, however, is whether the Panellit protection procedures are adequate to guard against the hazards of an accidental flow reduction to a tube operating in this manner. The result of a loss of a rear pigtail is an increase in coolant flow rate and, consequently, a reduced Panellit pressure. Furthermore, with the rear pigtail missing, the coolant flow no longer discharges through the Parker fitting on the rear header. Critical flow through this Parker fitting strongly influences the course of events following an accidental flow reduction to a normal ``O`` geometry tube. While it was strongly believed that the concepts developed and the Panellit procedures devised for a normal ``O`` geometry tube would apply to the case of a tube with a missing rear pigtail, evidence could only be obtained by transient experimentation. This document …

Studies were made to determine the stress equations necessary for the design and analysis of the OMR Core Vessel Support Structure. A parameter study of these equations applicable to the present core vessel design was also made. (W.L.H.)

A study of the solubility limits of iron and tin in alpha zirconium at temperatares below 500 deg C is in progress. Dilute alloys were prepared by arc- melting using the highest purity materials. Heat treatment of the alloys involves homogenizntion at 800 deg C followed by severe cold work prior to extended anneals at temperatures between 200 and 500 deg C in 50-degree intervals. Metallographic examination of annealed and quenched structures is the primary mode of investigation. Some magnetic susceptibility studies were also made. (auth)

ABS>Curves are given of pH versus percent extraction for Ac, Be, Bi, Ce, Dy, Eu, Gd, Hf, Ho, In, Fe, La, Pb, Nd, Np, Pu, Pr, Pm, Ra, Sm, Sc, Tl, Tm, Th, U, V, and Y by thenoyl trifluoroacetone in benzene. (T.R.H.)

One of the limitations of the Zircex Process is the large percentage loss of U which is carried off with the volatile ZrCl/sub 4/. Since the ratio of U to Zr in the fuel element is quite small (< 1%), a large percentage of U loss can be caused by small amounts of U in the ZrCl/sub 4/ sublimate. The U loss must be kept below 1% if the Zircex Process is to be adapted to industrial scale processing. (W.L.H.)