A study was made to determine the structural and functional adequacy of the EGCR graphite core design. Maximum stress and distortion of the core and the probable operating time before graphite cracking were determined. The major cause of stress is non-uniform fast-neutron flux, which causes non-uniform shrinkage of graphite components. The critical stress was found to be tensile. The criterion for determining the time of cracking of graphite columns appears to be the point at which the creep strain results in rupture. Column life before cracking may also be related to the maximum allowable stress at rupture. It was concluded that non-uniform shrinkage is the major cause of distortion of graphite components of the core. In general the amount by which a column tends to bow during 20-year core life exceeds the available free space. The combined internally generated and bowing stresses result in an average period before cracking of all core columns of 14 years, with a minimum period of 6 years. (M.C.G.)

To permit the irradiation of one dummy fuel element assembly for one operating period and to permit, during a subsequent operating period, the irradiation of one defected, four-rod-cluster UO{sub 2} fuel element assembly, in a KE front-to-rear test hole. The fuel material is natural UO{sub 2} of 95 per cent theoretical density; the cladding is zircaloy. The defect in the assembly is artificial and will be made before irradiation by drilling a .005in. diameter hole through the cladding near the mid-point of two of the rods.

The current status of the NPR fuel fabrication and testing program has been reviewed; suggestions have been made concerning areas requiring more information. This report is divided into a series of sections, each of which is a complete entity so a specific problem area can be considered apart from the entire NPR Fuel process. The final section incorporating conclusions and recommendations reaffirms the factors of major importance as well as evaluating those areas which are interrelated. Process problems have been considered from the aspect of manufacturing, not engineering.

This report summarizes investigations taken between October 1, 1959, to September 30, 1960 with the objectives of developing high-tungsten alloys which are readily fabricable into massive and complex shapes and which possess adequate strength for operation at temperatures up to 2000 F. Within this report, studies are divided into two main categories: alloy development and material fabrication techniques

The objective of this test is to determine the DR-Reactor effluent systems characteristics under 95 degrees Celsius bulk temperature operation. This proposed bulk temperature increase from 93.5 to 95 degrees represents a 33% decrease in the bulk temperature suppression below the boiling point. A major aim of this test will be to evaluate the degree of increased maintenance at this higher temperature operation. The basis and justification, test preparation and instrumentation, procedure, costs, outage time, hazards, standards, and responsibilities are discussed in this document.

None

Supersedes ITR-1421. Twelve large-diameter buried conduit sections of various shapes were tested in the 60- to l49-psi overpressure region of Burst Priscilla to make an empirical determination of the degree of personnel protection afforded by commercially available steel and concrete conduits at depths of burial of 5, 7.5, and 10 feet below grade. Essentially, it was desired to assure that Repartment of Defense Class I, 100psi and comparable radiations, and Class II, 50-psi and comparable radiations, protection is afforded by use of such conduits of various configurations. Measurements were made of free-field overpressure at the ground surface above the structure; pressure inside the structures; acceleration of each structure; deflection of each structure; dust inside each structure; fragmentary missiles inside the concrete structures; and gamma and neutron radiation dose inside each structure. All buried conduit sections tested provided adequate Class I protection for the conditions under which the conduits were tested. Standard 8-foot concrete sewer pipe withstood 126-psi overpressure without significant damage, minor tension cracks observed; standard 10-gage corrugated-steel 8-foot circular conduit sections withstood 126- psi overpressure without significant damage; and standard 10-gage corrugated- steel cattle-pass conduits withstood 149-psi overpressure without significant damage. Durations of positive pressure were from 206 to …

A technical and economic evaluation of a mercury-cooled fast breeder reactor is presented. The objectives of the program were to establish the technical feasibility of a fast breeder reactor cooled with boiling mercury and to evaluate the long-range potential of such a reactor power plant for production of economic power. Details of the conceptual design of a 100-Mw(e) reactor and system are discussed. The power cost from a mercury cooled fast breeder reactor was estimated as 21.4 mills/kwh which is competitive with the power cost for the initial Enrico Fermi plant. It was concluded that this reactor concept is technically feasible and has promising long-range economic potential. (M.C.G.)

A controlled radiation experiment was performed to determine the stability of fuel containing 28 wt.% UN dispersed in and clad with Type 318 stainless steel as compared with fuel containing 30 wt.% UO/sub 2/ dispersed in the same material. The specimens were prepared by hot rolling the fuel coupons in Type 318 stainless steel using the picture-frame technique for initial bonding and reduction. Final dimensions were obtained by cold rolling. A special radiation capsule was designed which contained heat control and enough thermocouples to ensure a good continuoustemperature history throughout the test. This capsule and the method by which the specimens were loaded are discussed in some detail. Because of the capsule instrumentatton. the known MTR position into which the capsule was placed. and the dostmeters placed tn the capsule it was possible to obtain a complete flux and temperature history of the capsule during the irradiation. When it was estimated that the specimen burnup was about 7.2 at.% of uranium-235 the capsule was removed from the reactor and returned to the Battelle Hot-Cell Facility. The postirradiation examination which consisted of fission-gas analysis, density and dimensional determinations, radiochemical and isotropic burnup analysis, and metallography is discussed completely in the report. …

A single swelling capsule was irradiated in DIR reactor as a test of the operation and design of a series of similar capsules to be charged later in a large scale uranium swelling experiment. The ratio of measured to calculated uranium temperature increased from 0.7 during the first two-thirds of the irradiation to 1.08 during the last one-third. Most of this change can be accounted for by a flux shift toward the rear of the reactor during the last one-third of the irradiation. Because of this flux shift it was impossible to determine an accurate temperature correction factor to be applied to the full scale test to follow. Radiometallurgy revealed an asymmetric radial temperature distribution in the fuel. Capsule tests showed that this asymmetry was caused by in complete NaK coverage of the rod; Capsule design change will prevent this.

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

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

In-reactor tests of self supported fuel elements is ribless process tubes have shown a significant reduction in hot spot failure incidents. It is believed that the support rails prevent gross fuel element cocking or misalignment which would allow inadequate cooling of a portion of the fuel element. It is reasoned that if similar spacing devices were attached to fuel elements for use in present ribbed tubes an appreciable reduction in hot spot failures would result. The problem then remains to select suitable spacing devices hereafter called called ``bumpers`` and to assess the increased energy losses associated with their use. Efforts to predict the energy losses which may be caused by fuel element support rails on bumpers have been somewhat discouraging. The odd shape of the support rails which appear something like a close-coupled suitcase handle preclude the rigorous use of available drag coefficients or contraction expansion coefficients. Hence it is necessary to make pressure drop measurements with a fuel elements design (size) which is considered to be quite close to meeting the actual pressure flow requirements. Then the final design is determined in view of these data. The data of this report were obtained in the 189-D Hydraulics Laboratory to …

Daily cash report for Imperial Sugar Company including a list of large withdrawals for the months surrounding the date showing the daily expenses and estimated balance for each entry. It also lists scheduled payments for bank loans.

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.

An ex-reactor test stimulating the charging of a tubular fuel element in a Zircaloy-2 process tube showed a problem of severe process tube scratching and galling may occur if metal sliding contact fuel element supports are used. Both the fuel supports and the process tubes were autoclaved. Although the conditions of this test were severe in that the sharp edges rather than a flat surface of the supports were in contact with the process tube, it pointed out the potential problem of galling by point on small area contact supports on heavy fuel elements. This difficulty had not been observed with the lighter tubular or cluster fuel elements used in KER irradiation. In addition to the loss of metal and thinning of the process tube after a number of charge-discharge operations, these scratches may lead to localized corrosion or act as points of stress concentration.

In recognition of the need for an integrated system plan for future development requirements for radiological instrumentation, this information report was written to provide the initial step concerning such a planning. The report is to be considered a basic information analysis of the present in-use plant radiological instrumentation, of the instrumentation now undergoing active development, and of the future development work contemplated.

This document details activities of the irradiation processing department during the month of November, 1960. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operation; 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 August, 1958. 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; and Financial Operation.

Report issued by the Brookhaven National Laboratory discussing fuel reactor test loops. The "design, construction, operation, and experimental results" (p. 1) are presented. This report includes tables, illustrations, and photographs.

Conclusions from the study: "Radioactive dusts are carcinogenic, and can cause cancer of the lung. complete dose response curves have not yet been determined. All the parameters that render this demonstrably toxic material (radioactive dust) have not yet been evaluated. It is strongly suggested by the experimental data that duration of radiological insult to the lung is an important factor in eliciting lung cancer. The atmospheric tolerance concentrations now in use seem to afford little margin of safety."

Until recently, bond adherence between the Al-Si braze layer and the uranium core was estimated by a method known as the ``chisel test.`` Fuel element jackets were removed from the core using a hammer and chisel, and the relative bond strength was estimated on a comparison basis. Bonds were classified as normal, below normal, or better than normal, depending upon the ease with which the jacket was removed. The weakest bonds were usually found one-to-two inches below the fuel element cap, but on occasions extended as far as five inches below the cap prior to making process changes in the duplex canning bath to improve bonding. After duplex furnace conditions were changed (silicon concentration, core submersion depth, and temperature) during September 1958, the weak bond area was reduced to a band 1/4 inch to 1 inch below the cap end of the piece. Recently, the stud-pull method for determining bond strength was developed providing a better method for measuring bond quality. Bond strength measured by the stud-pull method provides actual tensile strength measurements, which are valuable for test purposes and detecting in-process variations. This report summarizes tests made using the Instron Tensile Strength Machine for determining fuel element bond strength.

An investigation was conducted to establish an operating procedure for flushing water through a purification system demineralizer which was out of service for an extended period, and to determine the demineralizer serviceability. (J.R.D.)

None