ON NUCLEAR PHYSICS LABORATORY AT CANEL. The critical experiment facility at CANEL is described. Information of the mature of experimental assemblies and operations is included. Safety features of the building, equipment, and operations are pointed out. Possible accidents and the resulting hazards to surrounding areas are analyzed. The make-up of the surrounding area is described. (M.C.G.)

The operational characteristics of the Cosmotron were reviewed in brief. Since a rather complete account of this machine is now available in the September 1953 issue of Rev. Sci. Inst., this summary will include only the more recent utilization of the machine as a research instrument.

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A common and useful assumption is that the front-to-rear power distribution is a shortened cosine curve. This fact led to the calculation of slug factors'' for the longer charges to be used at K-Pile. It is hoped that the publication of these numbers will save time and prevent duplication of effort by those concerned in any way with slug power, surface temperatures, reactivity effects, etc. For the sake of completeness, slug factors for the older piles are included. Another useful ideal is that tube power distributions can be approximated by the assumption of a cylindrical pile and a cosine distribution of power outside of a central flattened region. This led to another double set of numbers since the values for the 2004 tube piles were again included for completeness. 2 figs., 6 tabs.

This document summarizes the amounts of radioactive contamination discharged to the ground from separation facilities through June 1956. Detailed data for individual disposal sites is presented on a month to month basis for the period July 1955 through June 1956. The major disposal sites in separation facilities, total volume of waste discharged at each location, and the gross amounts of plutonium and fission products discharged to the ground since startup are listed. This same data are presented on a month to month basis and also include information on the source of the waste stream and the settling facility used. Isotopic data are included for all disposal sites from which the waste was analyzed for specific contaminants. Estimates of contamination and volumes discharged to the swamps are also included.

In view of contamination difficulties within the Redox operation, number of items are being considered to improve this situation. It is the objective of this document to list and describe seven times which it is felt would contribute most toward improvement. It is also an objective to present for RDS-D-12 Group consideration the recommendations of representatives from Manufacturing, Technical, and Design in these matters with the expectation that a project proposal written in general terms for overall contamination improvement would be prepared.

The purpose of this course is two-fold. It should serve as a review for the metallurgist of the classical metallurgical concepts applied to such metals as uranium, thorium, plutonium, and zirconium. These metals are relatively unfamiliar to the metallurgist, but the concepts are still the same ones applied to familiar metals. The second purpose is to acquaint the non-metallurgist with the reasons for selection, advantages, and disadvantages of the various fuels and structural materials used in a reactor. Thus, there are two purposes: the first is conceptual, and the second is explanatory. A blending of the two must inevitably result in an over-simplification of the metallurgical concepts and an assumption of certain background material which, actually, may not be available to the non-metallurgist. The present lecture is an introduction to the material to be covered. This lecture will have two purposes. The material to be presented in the remaining lectures will be outlined briefly, and the various concepts to be covered in this course will be discussed.

Power rate of rise measuring instrumentation is being procured for all of the Hanford piles, and a prototype installation is now in service at D reactor. This instrumentation is expected to provide valuable assistance to the pile operator during the start-up rise to operating power; in order to best utilize the instrument, procedures and limits must be determined on the basis of the relationship between the flux rate of change and the instrument response. As with any measuring instrument, there is an inherent delay in the power rate meter circuitry; in addition there are greater delays associated with the beat capacity of the metal and water and the transit time of the cooling water. Furthermore, reactivity changes from rod withdrawal and metal coefficient draping are not instantaneous, and the flux is not a simple function of the reactivity changes. Because of the time lags involved the rate meter response cannot be identical to the actual flux rate of change; however, an exact solution for this response in terms of all of the variables would be inordinately complex. The purpose of this study is to show approximately the changes in power level or rate which might occur in practice relative to …

There were twenty slug jacket failures during the month of January 1952. Of these, fourteen were end cap failures, five were split slugs and one has not been removed. A total of 311.4 hours of outage time was required for removal of these ruptured slugs from the reactors. The detection, removal and radiation aspects, along with the slug data, are shown on the attached sheets.

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Six 1.6% enriched Zircaloy-2 jacketed seven-rod cluster elements with modified end supports and one Zircaloy-2 jacketed Pu-A1 seven-rod cluster element will be irradiated in KER Loop 1 to an exposure of about 4500 MWD/T.

In a preceding lecture all the components necessary to secure a controlled nuclear chain reaction were discussed. The business at Hanford is to use these components in the most efficient manner possible to manufacture the desired products of irradiation - the primary product being Pu. In this lecture of the variables of such operation are considered.

With higher power levels and probably higher river temperatures, the forthcoming warm water season is expected to impose bulk outlet temperature limits, with consequent production losses, on all Hanford reactors to an even greater extent than last year. The problem has received the attention of a number of management and engineering personnel, but the basis for this study was a letter originating in the Production Operation suggesting two alternate schemes for increasing the flow of process water to the rear face piping, thereby reducing bulk temperatures for a given power level. The purpose of the suggestions was to provide an interim solution to the problem pending I & E loadings. Loading with I & E slugs will increase reactor flows to some degree, depending on the actual slug dimensions used, and will probably result in encountering other limits than the B.O.T. limit. The applicable period for study was therefore taken to be this summer only, since I & E loadings are expected to begin in June or July at 105-D and during the fall at 105-F. The Managers, D and F Reactor Operations, requested Plant and Industrial Engineering Operation to investigate and compare the feasibility and applicability of the suggested …

This report presents the cold shakedown run for the UO{sub 3} plan. One complete batch conversion will be made in each of the eighteen Calcining Pots using unirradiated uranium food solution. The objectives of this series of conversions are as follows. A mechanical test of the conversion units; a shakedown of the unloading, pulverizing and material handling equipment; a functional test of the Nitro Acid Absorber system operating at approximately 5 to 20% of its design capacity; the determination of the optimum temperature and time cycle conditions for 60% UNH feed solution. These conditions may be based upon calcination of the initial charge of 60% UNH solution to form UO{sub 3}, in one pot; a confirmation of the calibration of instruments; and to provide operating experience and process know-how.

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Design and construction schedules and forecast monetary allowances for projects in the 100 Areas, 200 Areas, 300 Areas, and 700 Area of Savannah River Plant have recently been reviewed. There has been no change in the overall funds now forecast for the program, although there has been changes in the distribution of the total between various areas.

The objectives of this test are to obtain a current knowledge of the H Reactor flow system characteristics under various transient conditions which have a reasonable probability of occurrence. The test data will be used to revaluate: (1) the H Reactor bulk effluent temperature limits, (2) the margin between the Ball 3X riser pressure trip settings, and (3) the quantitative adequacy of emergency flow available from the combined high tanks export system. This test will consist of two parts: (a) Measurement of the flow characteristics during transition from the primary coolant system to the secondary coolant system (190 steam driven pumps.) (b) Measurement of the adequacy of the last-ditch emergency water system (combined high tanks-export system) during a simulated primary and secondary coolant system failure.

The purpose of this report is to summarize all actual and potential power level limits under which the K Reactors may be forced to operate during the next year. In addition, a summary of the methods being followed to eliminate each is to be discussed along with a graphical presentation of the relationship of the various limits. For the purpose of this report, it was assumed that all power levels will be attainable from a physics aspect since these problems are discussed elsewhere. It is not the intent of this report to recommend that these power levels be attained but only to indicate what production levels are possible through operation on any one limit. Seven outlet temperature limits govern operation at the K Reactors. They are: TBI Limit; Crosstie Limit; Rear Header Saturation Limit; Bulk Outlet Temperature Limit; Tube Corrosion Limit; Temperature Monitor Limit; and Rupture Control Point. Operation during the summer months of 1957 will require cuts in power in order to remain within these limits unless modes of circumventing a few are found. The limits are shown graphically in the appendix at the various flow rates probable. From these curves it is evident that the Low Trip TBI …

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Simultaneous reduction of coolant to several or all reactor tubes raises concern not only for the adequacy of protection in the individual process tube but also the reactor as a whole. In event of such flow reduction, the heat generation does not decrease until at least 1.4 seconds have elapsed following the accident. Thus, the water temperature from each tube will rise, and result in an increase in the bulk water temperature. If the increase in bulk water temperature is such that saturation temperature at the top of downcomer is reached, pressurization may occur at that point and exceed the maximum recommended working pressure limit (approximately 1 to 2 psig). The purpose of this report is to present experimental data on a series of tests which were made to simulate flow reductions to a K type process tube by simulated front header pressure decreases.

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The 234-5 Building ventilation exhaust is continuously sampled for the purpose of estimating the amount of radioactive (alpha emitting) material discharged to the atmosphere. Although a record is kept of the gross amount of radioactive material discharged, few data are available concerning the size and kind of active particles in the exhaust air. Knowledge of the particle size permits: (1) an estimate of the validity of samples drawn through the sampling system, (2) a better knowledge of what the active particle ground deposition pattern might be, and (3) may provide information relating to filter performance. The kind of radioactive material discharged is important in determining relative health hazards. The object of this work was to determine the size and kind of radioactive particles in the 234-5 Building ventilation exhaust. A secondary objective was to review present routine sampling of the stream with particular regard to the particulates to be sampled.

This test consisted of a series of helium purges of the pile gas atmosphere to bring the helium content up to about 70 to 80% helium followed by a series of CO{sub 2} purges to restore the pile atmosphere to its initial composition. The purges where spaced roughly four hours apart to enable the reactor to reach a steady temperature before purge was undertaken. The test results are presented in this report.