In order to investigate the rupture stability of uranium elements irradiated at power generations per unit length larger than those encountered in natural uranium fuel elements, three partial columns of 1.44% U{sup 235} enriched, internally-externally cooled, uranium elements were irradiated in PT-IP-1-A. After discharge and examination of the fuel elements 25 pieces (182.5 pounds) of this metal were available for special analysis of U{sup 235} burnout and plutonium content. The average exposure of these pieces was 2,187 {+-} 6% MWD/Ton. The purpose of this document is to summarize some calculations of buildup of plutonium isotopes and burnout of U{sup 235} in an attempt to correlate calculations with the results from experimental analysis.

Summarized herein is the available information on the subject, ``Reactor Gamma Flux Monitors.`` Apparently, a misunderstanding exists between the U.K. representatives and the related information presented by Dr. Nilson at the Argonne meeting July 24 and 25th. Dr. Nilson`s reference at the US/UK meeting was to the use of gamma-compensated neutron detection chambers at Hanford. Such chambers have been designed and used in one old reactor for a short period as a prototype for a detector for the octant monitor system. No gamma compensated chambers are in any of the reactors at present. Under development is a neutron flux indicating system which will operate in the intermediate range (10{sup {minus}7} to 10{sup {minus}1} full power) and will use commercially available gamma compensated ion chambers. These will be used to provide signals for reactor period meters. At Hanford there has been no operating experience with gamma flux monitors as such or with extensive in core neutron flux monitors. Characteristics of systems under development are given for information as well as current information on the octant system.

The feasibility of adding four ``corner`` rods to the K Reactor`s 20-rod HCR system is being investigated. It is necessary to compare the costs and production gains of this method for gaining additional reactivity and heat distribution control with other supplementary control systems. This document presents the results of an investigation of the production gains that could be made by adding four rods to the KE Reactor HCR system.

The use of organic cooling media for nuclear reactors operating at high power levels predicates the use of a coolant which will not react violently with metallic uranium in the event of a fuel element failure. This report describes the testing, and subsequent examination, of two pieces of irradiated uranium which were immersed in monoisopropylbiphenyl (MIPB) at high temperatures and pressures for periods of time up to twenty-five days. The uranium samples had different irradiation histories and cooling times. Similar experiments had been performed with unirradiated uranium by the Corrosion and Coatings Operation, and it was wished to determine whether irradiated uranium would react with MIPB in a different manner.

The purpose of this work has been to determine the production rate of tritium in fringe Li-Al alloy columns with the degree of precision necessary for economic analyses of such reactor loadings. These results are provided for use in such an analysis. This experiment indicates the production rate of tritium in the outermost fringe tubes to be T = 0.0216 M{sub E} = 0.175 M{sub t} where T = grams of tritium per full length (67 pieces) charge of Li-Al alloy material; M{sub E} = MWD/adjacent ton of E metal; M{sub t} = MWD/adjacent tube of E metal. The above values should apply for fringe loads utilizing greater or smaller quantities of E metal; that is, for isotope production loadings which are over or under-compensated from a reactivity standpoint. In the actual test load it was calculated that one gram of tritium and 13.5 grams of Pu were made for each 21.3 grams of U-235 burned up. During the same time interval the displaced uranium loading would have generated 24.3 grams of Pu and burned up 29.9 grams of U-235. The factor which seems to limit the accuracy with which these data can be interpreted is the ratio of the …

This document provides the conclusions and recommendations of the study of purging of surface condenser tubes. (FI)

Graphite shrinkage of 0.60 inches (approximately 30 per cent) has been observed at F Reactor in the region of maximum graphite distortion. A reduction of graphite distortion at B Reactor of 0.60 inches in the top-center and 0.32 inches in the top-far region also has been observed. Analysis of the vertical traverse data indicates that continued graphite annealing with resultant decrease in graphite distortion might be anticipated although perhaps at a reduced rate. No significant detrimental effects have been observed from operation with either the increased charge length or the flux slightly skewed to the front of the reactor. Operation with severe flux skewing resulted in detrimental control effects. It is recommended that operation with the flux skewed to the front of B and F Reactors through the use of a longer charge (centered slightly upstream of the graphite century-line) be continued subject to periodic review.

In the course of trying to predict the shape of NPR horizontal rod and process tube channels after several years of operation, data from 105-K were referred to. While looking at the 105-K data the following possible explanation of the ``K`` reactor distortion phenomenon occurred to me. It is written up in this document as a possible aid to those responsible for evaluating the stack distortion problem in the ``K`` reactors. Data and references to data are presented first and then an analysis of the data is presented. The author does not purport to be an expert on the ``K`` stack distortion or graphite technology but has written this analysis because it appears to be a plausible explanation of existing conditions.

Beginning in February, 1958, a sufficient quantity of seven-inch dip canned I & E depleted uranium fuel elements was prepared for irradiation to produce eleven kilograms of plutonium, containing at least twenty per cent of the Pu-240 isotope, as authorized by the Atomic Energy Commission. Subsequently, eighty-four columns in C reactor were partially charged with the finished depleted fuel under PT-IP-132-AC. To date, ten depleted ruptures have been sustained, after being irradiated six to eight months toward a planned accoumulated goal exposure of 210 MWD per column or a total irradiation time approximating 12--14 months. The mechanism and cause of these failures is being thoroughly investigated. This document summarizes the fabrication history, irradiation experience to date, rupture examinations, and an investigation of process conditions which may have contributed to the high incidence of ruptures.

This is the monthly report for the Hanford Laboratories Operation. Metallurgy, reactor fuels, physics and instrumentation, reactor technology, chemistry, separation processes, biology, financial activities, employee relations, laboratories auxiliaries, radiation protection, operation research, inventions, visits, and personnel status are discussed. This report is for October 1958.

The July, 1958 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)

The distribution of the heated effluents discharged by Hanford reactors to the Columbia River has been a matter of interest since the early design stage of the first reactors. The pattern of this distribution is a major factor in determining the extent to which a downstream reactor is affected by those upstream, as well as the localized effects on the ecology of the river. Pollutional characteristics of the effluents are three - heat load (or temperature increase), chemical contents and radioactivity. The latter has received the greatest attention in connection with potential personnel exposure and effects on river biota; it has been assumed however, and generally confirmed by sampling that the measure of distribution of any one of these characteristics in the saw an for the others. Observed distributions of radioactivity for various river and reactor flow rates are documented. Unfortunately, any extrapolation of those observed distributions to altered flow conditions of river regimes is of questionable validity. Mathematical models of the problem have been formulated but have been of little value due to the necessity of measuring certain parameters under the conditions for which a solution is desired. Even so, calculated distributions provide only general patterns and would not …

This report summarizes the design and development work completed since May, 1957; the current status of fuel element matching problems, and the plans for future modifications to improve machining techniques.

The reliability of the K-Reactor water plants has been reviewed by Research and Engineering personnel; this study augments a brief analysis of the KW generator failures incident given in my letter of October 6, 1958. Our attention has been directed largely to ascertaining the requirements of the system, the consequences of possible component failures, giving general assessment of the primary features upon which reliability depends and identifying potential improvements which would increase reliability or moderate the consequences of a major failure. Our attention is restricted to failures which might occur as a result of equipment failure, operator error or certain natural causes such as earthquakes; we have not weighed special considerations brought on by enemy action such as sabotage, for example, in this evaluation. R. S. Bell is providing an analysis of the operating and maintenance aspects of K water plant reliability to complement this engineering study. It is concluded that the basic elements of the system lend themselves to reliable operation. The lines of defense appear adequate in depth, that is, a primary system operated with BPA power, an emergency ``backup`` system consisting of two essentially independent steam driven generators driving selected pumps, and a ``last ditch`` system providing …

The purpose of this report is to present a set of standard pressure-drop curves for various fuel elements in process tubes of Hanford reactors. The flow and pressures within a process tube assembly under normal conditions are dependent to a large extent on the magnitude of the pressure drop across the fuel elements. The knowledge of this pressure drop is important in determination of existing thermal conditions within the process tubes and in predicting conditions for new fuel element designs or changes in operating conditions. The pressure-flow relations for the different Hanford fuel element-process tube assemblies have all been determined at one time or another in the 189-D Hydraulics Laboratory but the data had never been collected into a single report. Such a report is presented now in the interest of establishing a set of ``standard curves`` as determined by laboratory investigations. It must be recognized that the pressure drops of fuel elements in actual process tubes in the reactors may be slightly different than those reported here. The data presented here were obtained in new process tubes while reactor process tubes are usually either corroded or filmed, depending on their past history.

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 …

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The test detailed in this report is a part of a program seeking ways to reduce the amount of radioactivity in reactor effluent water. Arsenic-76 is a major contributor to the gastrointestinal, radiation dose received by downstream users of river water. Phosphorus-32 is the principal radioisotope in the flesh of whitefish caught in the vicinity of the Hanford reservation. Recent activation analyses of the commercial grade sulfuric acid used in process water treatment have indicated that this acid may be an important source of the parent elements that are transformed to arsenic-76 and phosphorus-32 in the reactors. The purpose of this test is to determine if the impurities in commercial grade sulfuric acid have a significant effect on the release rates of arsenic-76 and phosphorus-32 to the river in reactor effluent water.

The composition of dissolver off-gas was determined for seven different dissolvings of uranium in nitric acid (both with one and two dissolvers in operation) at irregular intervals during February, March, and April, 1958. Samples were taken at the Purex 293-A facility absorber inlet, scrubber inlet, and scrubber discharge. The absorber was operated with and without reflux, and the scrubber was operated with water or with approximately 10% sodium hydroxide. This memorandum describes the sampling procedures and analytical methods and gives the gas analyses found.

Recent observations and measurements of the K Reactor stacks have revealed that horizontal shifting of the graphite has occurred. This shifting has caused the stack sides to bow outward as much as three inches. In addition, gaps large enough to admit 3X balls have been observed between blocks in certain regions of the stack. The term ``distortion`` as used in this report refers to this horizontal shifting and the resulting formation of gaps. The vertical growth and contraction that have been observed are primarily exposure phenomena and are similar to the radiation damage experienced at other HAPO reactors. Considerable data have been accumulated in an attempt to define the magnitude and patterns of the distortion, its cause or causes and the effects it will have upon reactor operation. This report compiles the known data and discusses what is known or suspected about the causes and effects of the distortion. As additional data and information become available, supplements to this report will be issued.

The February, 1958 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)

It has been noted on occasion that the side rupture rate at lower exposure is higher than one would expect in order to be consistent with the mathematical model (Weibull curve) used describe the rupture data. This has apparently been due to a few ruptures occurring at exposures lower than one would anticipate on the basis of the model. Restricting themselves to fuel elements irradiated under similar conditions of power and temperature, it is difficult to assert that these ruptures are anomalous with respect to the model used since there are so few ruptures involved. For this reason, some means of combining data was called for in order to see whether or not at least some of these low exposure ruptures were inconsistent with the model, and, if so, to revise the model in order to better describe side rupture performance. In this report, the results of an analysis of side failure data for failures covered in a previous report are given. Also included is a discussion of possible implications derived from recognition of the existence of this revised model.

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Brittle bonds and the factors controlling their formation have been of substantial concern in the production of dip canned fuel elements. Detection of brittle bonds has been by the chisel test and by metallographic examinations. At best, these are qualitative tests and do not establish the degree of brittleness. For this reason bond tensile strength analysis has been suggested. Tests have been run to determine if a change in canning variables could be detected by a change in bond strength.