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.)

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 …

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 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.

This report investigates a study of the accuracy of the power level calculations. Some characteristics of this calculation can be adduced but the accuracy in the sense of freedom from bias leads to an impasse. It presumes that there is another measurement for comparison purposes for which there are engineering reasons to believe, the inherent error is smaller. For the power level values as calculated by the Foxboro Power Level Calculator no such measurement or calculation exits. However, the calculations can be studied to determine whether it is worth-while to make further refinements relative to the various functions of the power level calculation. The accuracy requirements vary according to use where the most stringent requirements are in the control of SS Materials.

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 …

Tests in which aluminum-jacketed, Al-Si bonded uranium fuel elements were baked at various temperatures have shown there is a time-temperature relationship for Al-Si layer decomposition. For heat transfer and secondary coolant barrier considerations, the extent of bonding layer deterioration during fuel element irradiation is important. Under present reactor operating conditions, Al-Si bonded fuel elements show evidence of internal bond deterioration, and to a lesser degree, external bond deterioration following irradiation. Such evidence has aroused concern for the ability of the Al-Si bonding layer to withstand future reactor operating conditions. Currently, several potential uranium fabrication and canning process improvements are being developed to further advance fuel element stability and performance. Optimization of process conditions based on these improvements may provide the necessary Margin of safety for good bond layer integrity during irradiation. Before a decision can be made to continue improvement of the present process or convert to a new canning process, more information on the stability of the present fuel element bond is needed. The purpose of this report is to summarize data derived from past and present testing and to recommend courses of action to more fully evaluate Al-Si bond integrity under anticipated future reactor operating conditions.

The fixation of radioactive waste in a stable solid media by means of calcination of these aqueous solutions has been the subject of considerable-effort throughout the U. S. Atomic Energy Commission and by atomic energy organizations in other countries. Several methods of doing this on a continuous or semi-continuous basis have been devised, and a fev have been demonstrated to be feasible for the handling of non-radioactive, or low-activity, simulated wastes. Notable among methods currently under development are: (a) batch-operated pot calcination of waste generated from reprocessing stainless steel clad fuel elements (Darex process) and Purex waste, (b) combination rotary kiln and ball mill calcination of aluminum nitrate (TBP-25 and Redox process), and (c) fluidized bed calcination of TBP-25 and Purex wastes. Although a considerable amount of engineering experience has been obtained on the calcination of dissolved salts in a fluidized bed, and the other methods have been the subjects of a great deal of study, none of them have been developed to-the extent which would rule out the desirability of further investigation of other possible methods of calcination.

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 June 1960. (FI)

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.

The reactivity control capacity of the N-reactor control systems has been calculated by methods which take into account the absorption of epithermal as well as thermo neutrons. Earlier calculations were made using a method which is satisfactory in control calculations for the existing Hanford reactors but which largely neglects the epithermal absorption. The new calculations were undertaken because of recent evidence that the epithermal absorption is more important to the NPR control strength and because of some concern regarding the margin between previously calculated control strength and anticipated control requirements. The calculational methods used in the present study and the resulting calculated control capacities are reported in this document.

To permit the irradiation of one dummy fuel 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.

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.

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

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.)

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Summary statements are given of significant findings related to the distribution, characteristics, and biological availability of fall-out debris originating from testing programs at the Nevada Test Site during the past decade. The delineation of fall-out patterns has been accomplished by the use of aerial and ground monitoring surveys. Only about 25% of the total amount of fission products produced by tower-supported detonations were deposited within distances corresponding to fall-out time of H + 12 hr; a much smaller quantity was deposited by balloon-supported detonations. Most of the fall-out debris that was redistributed by various environmental factors after original deposition consisted of particles < 44 ..mu.. in diameter; the particles in this size range also represented the predominant contamination on plant foliage. /sup 90/Si levels in surface soil ranged from 31.9 to 142 mc/sq mile in virgin areas near known fall-out pattern midlines and from 7.5 to 22.7 mc/sq mile in agricultural areas that in some cases did not coincide with fall-out pattern midlines. The accumulation of radioiodine by native animals was observed to be a function of distance from GZ. /sup 140/Ba, /sup 91/Y, /sup 89/Sr and /sup 90/Sr were major bone contaminants. Post-series sampling of native animals indicated that …

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.

The Project CGI-791, Reactor Confinement for Hanford production reactors, vas conceived as a result of deliberations concerning the release of the Wahluke Slope secondary exclusion zone for private use, and initiated by an authorization and directive from the Atomic Energy Commission to proceed with such a project on a high priority basis. This authorization was coupled with the establishment of limits on reactor heat-generation rates and fission product inventories. These limits were to be re-evaluated after significant progress has been made toward providing some undefined degree of control of the ultimate release to the environs of any air-borne radioactive fission products which might be released into the building housing any Hanford production reactor. It is the purpose of this report to summarize the philosophical and technical bases for the nature and degree of confinement provided under project CGI-791, to describe the project, and, further, to indicate the expected effectiveness of this confinement system. In discussing the expected effectiveness of the project this report is responsive to a request from the AEC-HOO.

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.

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Summary statements are given of significant findings related to the distribution characteristics, and biological availability of fall-out debris originating from testing programs at the Nevada Test Site during the past decade. The delineation of fall-out patterns has been accomplished by the use of aenial and ground monitoring surveys. Only about 25% of the total amount of fission products produced by tower-supported detonations were deposited within distances corresponding to fall-out time of H + 12 hr; a much smaller quantity was deposited by halloon-supported detonations. Fall-out particles less than 44 (For in diameter are presumed to be of the greatest biological significance. About 30% of the fall-out radioactivity from tower-supported detonations was contained in the 0 to 44 (For particles as compared to almost 70% for balloon-supported detonations. Fall-out debris from halloon- supported detonations was also much more water and acid soluble than was the debris from towel-supported detonations. The <44 (For fallout particles contained a higher percentage of Sr/sup 89/, Sr/sup 90/, Ru/sup 10/ / sup 3/, and Ru/sup 106/ than did larger sized particles. There was a higher percentage of these radioelements in the particles from balloon-supported detonations. Within distances corresponding to H + 12 hr fall-out time, balloon- …