We have successfully conducted a series of experiments involving scattering of high energy pions and protons from a target containing polarized protons. Results of some of these experiments were reported at this conference, and in the literature. Proton polarizations as high as 65% have been measured; the average polarization during sustained data-taking has been typically 45%.

It has been known for a hundred years that formaldehyde polymerizes to carbohydrate substances in alkaline media. Although the reaction has long attracted much attention, only recently has a detailed qualitative analysis of the products been carried out by chromatographic methods. We have started to re-examine this reaction by combining chromatography with radioactive tracer techniques in the hope of refining the quantitative aspects of the analysis. Our particular interest has been to develop methods for determining the relative proportions of ribose and ribulose in the mixtures of sugars formed in basic media, as well as under other polymerizing conditions. The finding of large amounts of these sugars might help to explain the occurrence of ribose as the only basic sugar in the fundamental replicating molecules--the nucleic acids. Formaldehyde is thought to have been present in the primitive reducing atmosphere which existed before life first appeared. The ribonucleic acids must have appeared in the constitution of reproducing systems at a very early stage in the development of living organisms. In this study, the polymerizations of formaldehyde were carried out in calcium hydroxide suspensions at 40{sup o}. Aliquots of the reaction mixtures were withdrawn at after various time intervals and the alkali …

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In order to evaluate entry into the atmosphere of an oblate (equatorial bulge) earth a standard atmosphere was used in conjunction with the standard rotating oblate spheroid. The density variation encountered in any one polar orbit around the earth will vary by about a factor of ten, due to oblateness. Therefore, the important effects on the re-entry trajectory are atmospheric density and oblateness of the earth. The SNAP configuration burn up characteristics will differ considerably between the steep (6/sup 0/) entry of the NASA Scout test and the orbital decay trajectory. However, the test can verify heat transfer rates on the actual configuration, and by proper calculation of material response, the actual decay breakup prediction can be improved.

Process development personnel have been seeking means of improving the (after thermal exposure) adhesion of the S8DR ceramic coating to chromized Hastelloy N tubing. This effort has led to the incorporation of an elevated temperature chromium diffusion treatment prior to the application of the ceramic coating. This report describes the results of tests performed to ascertain the influence of the additional thermal treatment on the 1400/sup 0/F creep-rupture properties of S8DR tubing.

This manual covers the technology involved in the 200 Area process for the recovery of Np/sup 237/ from certain aqueous waste streams in the separations plants, for the recovery of NP/sup 237/ and Pu/sup 238/ from irradiated NpO/sub 2/-Al slugs and for the fabrication of NpO/sub 2/-Al slugs. The manual contains sections on the fundamental chemistry, the primary recovery of Np by ion exchange, the decontamination of Np by ion exchange, the processing of NpO/sub 2/-Al targets, the separation and purification of Np/sup 237/ and Pu/sup 238/, the finishing of Np, the preparation of NpO/sub 2/, the disposal of spent resin, and the safety aspects of the handling of hydrazine. The section on the fabrication of NpO/sub 2/-Al slugs will be added later. 76 refs., 22 figs.

This report covers the activities with regard to on-site customer irradiations as of the above date. The report covers the status of materials undergoing irradiation, awaiting disposition, material shipped during the month, and current status of all reactor test holes. The integrated exposure values are calculated in accordance with HW-62781, {open_quotes}Status Report Equations.{close_quotes} The integrated exposure reported is the average exposure received by the surrounding uranium columns. Conversion to exposure received by the sample is left to the customer.

The results of a multi-group, diffusion type code computation for the fast neutron flux distribution in C, K, and N reactors are presented. Relative Ni activations in conjunction with the calculated neutron spectra are used to obtain values for the integrated fast flux per MWD/AT for C and K reactors. The relative, theoretical, integrated fast flux values at the same adjacent fuel powers in C, K, and N reactors in the mid-plane of the filler layers are: (1) C = 1.0; (2) K = 1.3; and (3) N = 2.6. For C reactor, the best estimate of the integrated fast flux per MWD/AT is: 5.0 {times} 10{sup 16} nvt (> 1 MeV) = 1 MWD/AT.

This report provides a compilation of memorandums and studies related to a Hanford production cut-back.

Recent borescoping of the KER-1 tube revealed several scratches, pits, and gall marks on the internal wall of the tube. These deformations could limit the operating temperature and pressure of KER-1. This report is a summary of operating history and is compiles to assist in determining what contributed to the condition of the tube.

In recent months, several process tubes installed during 1961 have been removed and examined out-of-pile in connection with Tube Corrosion Monitor Studies.1 Most of the tubes were believed to be uniform-wall .080in.-nominal-rib-height tubes. The ex-reactor dimensional measurements obtained near the inlet flange showed,that the tubes did not have uniform walls or .08011 nominal rib heights. Subsequently, dimensional measurements were obtained on approximately 200 new process tubes in an effort to determine the extent of dimensional variations in new process tubes as received from the vendors. The effects of the variations on tube corrosion and tube life are discussed in this report.

A research and development project for study of the effects of reactor effluent on Columbia River water quality is being sponsored by the AEC Division of Production. Work was started in October 1962, the first effort being aimed at furnishing an immediate answer to a security question: how closely can Hanford production be estimated by measurement of river temperatures? An interim progress report gave the results of the preliminary investigation. During the calendar year 1963, the study was expanded to meet broader program objectives. This document is a progress report for the year, covering the temperature and effluent distribution phases of the program. Progress on the chemical characteristics phase of the program is documented separately. Figure 1 shows the section of the river under study.

Americium (atomic weight 241) grows into plutonium as a result of the 13-year half-life beta decay of Pu{sup 241}. An appreciable quantity of Am{sup 241} has.grown into the scrap that accumulated following shut-down of the Recuplex facility, and which will be processed during the initial period of operation of the plutonium reclamation facility. Current interests in trans-plutonium research and in isotopic heat sources make it desirable to consider the recovery of this Am{sup 241}. The americium contained in scrap that is processed through the reclamation facility should appear almost quantitatively in the aqueous waste stream (CAW). Subsequent processing of the CAW through the DBBP extraction column (CW) of the waste treatment facility should result in further separation of plutonium and americium by extracting some of the plutonium while most of the americium remains in the aqueous waste stream (CWW). The fact that a low free nitric acid concentration will favor the extraction of americium into TBP-type solvents can be used to recover americium from the large volume, high salt concentration CWW stream. The flowsheets in this document present the chemical conditions for effecting this recovery.

Salt wastes from the Redox solvent extraction process have been routed to the 241-SX tank farm for storage since May 21, 1954. Tanks in this farm contain wastes from three types of irradiated uranium processing: (1) low (approximately 250 MWD/T) and (2) high (approximately 600 MWD/T) natural uranium; and (3) E-metal. It is estimated that approximately 1.6 Kgs of Am-Cm, 17 Kgs of Np-237, 20 Kgs of Pu and 8 tons of uranium are present in the sludges from wastes generated through 4-30-64. A total of 12.1 Kgs of the 29.1 Kgs of Np-237 estimated has been recovered or is present in the Redox Np-237 accumulation cycle as of 4-30-64. Because of the potential for unmeasured losses (such as Pu or U precipitates) the total plutonium and uranium may exceed the quantities stated herein. A breakdown of the estimates by tanks is provided.

Four years of intermittent development has culminated in the successful test drilling on-reactor, of one step plug hole. Since 1960, several different drive units and many diamond, carbide, and tool steel cutting heads were tried unsuccessfully in attempts to bore a stepped hole in a mockup of B, D, P side shielding. Success was finally achieved in 1963 using a standard horizontal boring mill and tool steel cutters. With slight modifications, this same equipment was successfully used in an on-reactor test drilling at F Reactor in December of that year. The on-reactor test revealed the need for improvements in the equipment. Chip flow was inadequate and appeared to be caused by poor air flow in the core receiver. Cast iron cutting technology used on the mockup did not work on-reactor and had to be revised on the spot. The graphite did not break up into chunks as desired and had to be manually removed. None of the steel cores were more radioactive than 50 mr/hour at approximately 2 inches, but the cast iron core read 300 mr/hour at 10 feet. It had to be handled very quickly. The dose rate for handling the graphite was 400 mr/hour. It, too, was …

Localized cladding instability or necking has resulted in failures of the Zr-2 cladding of rod cluster and tubular fuel elements and fuel rods in NaK filled capsules. The cause of this non-uniform cladding strain is thought to be a combination of the effects of irradiation and nonuniform cladding thickness. To determine the influence of these two factors and that of temperature on the susceptibility to failure, and to obtain data related to cladding thickness tolerances for the NPR fuel elements an irradiation experiment was designed. This experiment consisted of eighteen capsules each containing three Zr-2 clad uranium rods. Cladding thickness variations of up to 25 percent were introduced in the cladding by machining ``grooves`` longitudinally on the outer surface. The temperatures of the cladding were established by the thickness of the capsule wall. These capsules, charged in DR Reactor in October 1961, were prematurely discharged after five days irradiation because of a rupture. The cause of the rupture was established to be overheating and accelerated corrosion of the Zr-2 capsule body as a result of inadequate coolant to the capsule surface.

The high pressure loop installed in the 3X3 reflector position of the ETR and the associated instrumentation to detect and study failure mechanisms handled the rupture tests without difficulty. Failure of the elements was initiated by shearing off a projection on the fuel elements. The first test of the series used previously unirradiated seven-rod clusters. After the projection was sheared off the fuel elements were operated for seven hours with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second and third tests used fuel which had been irradiated to 2400 MWD/T at Hanford prior to insertion into the ETR. The second test was operated for 14 hours after the projection was sheared off--again with no failure. The third test was operated for only 33 minutes after the projection was sheared off before fission product activity in the loop water caused the test to be terminated.

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This report presents an analysis of rupture experience at C Reactor over the past two years. The purpose of the study was to provide a basis for revising, if necessary, the current metal performance level multipliers for C-II-N and C-II-E material.

C Reactor horizontal control rods were originally designed to have 32 feet of poison, made of 64 six inch ``cans`` each consisting of two concentric cylinders sealed at each end and the annular space between them filled with boron carbide powder. It was discovered before startup that under irradiation the neutron, alpha reaction in the boron could cause a pressure buildup and rupture of the sealed section. As an expediency cylinders wrapped with 72 miles thick cadmium metal were substituted for the boron ``cans`` and the pressure buildup problem was eliminated. However, since for a unit volume, natural cadmium contains fewer high cross-section nuclei than natural boron, the lifetime of one of these cadmium rods in Hanford flux levels is limited. Five of the original 15 cadmium rods were replaced in 1957 with boron rods of improved design. The primary purpose of this document is to present the results of a study to evaluate the extent of burnout in the remaining ten cadmium rods and their present rate of burnout so that replacement can be scheduled before these rods start losing significant reactivity poisoning effectiveness.

Solidification of intermediate-activity wastes is a major goal of the CPD Waste Management Program. Plans are to reduce the wastes, by evaporation, to salt cakes in existing tanks, thereby insuring safe, long-term storage of contained fission products regardless.of tank integrity. Initiation of these plans at an early date is necessary to offset the expected increase in tank failures and to provide space for future wastes. Major decisions of the program relate to selection of the evaporative method to be employed. The requirements of in-tank solidification were therefore reviewed to determine if the choice of evaporative systems can be made at this time. The relative potential of Bentube evaporation and submerged combustion for meeting these requirements were analyzed on the basis of available information, including actual performance of the Bentube facility at the Savannah River Plant (SRP).

Information resulting from recent graphite investigations in the K Reactors were presented to the Council. Separations as much as three to four inches were reported. Contraction of the graphite within the fueled regions of these reactors is now about one per cent in the direction transverse to the extrusion axis of the graphite and about two per cent in the parallel direction. Saturation is not expected until shrinkage of at least two to three per cent is observed in the transverse direction. The contraction is resulting in two principal operating problems; control rod entry and loss of 3{times} balls in the separations following a ball drop. The principal purpose of the review concerned only the ball system and particularly the philosophy governing the system`s use. Following a historical review of the ball system philosophy, our present views were developed by the consideration of six questions and answers. Our points extracted from these questions were (1) the IPD reactors require a backup safety control system (2) the backup system should be separate from the primary safety control system, (3) the backup system should be activated automatically upon accidents capable off disabling the safety rod system, (4) the system actuation should be …

Compensation for the positive long-term reactivity transient associated with Hanford reactor may be accomplished in two ways: The addition of a poisonous material (rods, splines, etc.) to the reactor, or cooling the moderator by changing the gas composition. The objective of this study is to investigate the reactivity and temperature effects and the associated operating problems if any, resulting from the use of nitrogen instead of carbon dioxide as a constituent of the reactor atmosphere.

Reactor water leaks cause outage time for locating and eliminating the leaks and for repairing damage paused by the vater that enters the reactor. The justification for equipment and process changes necessary for reducing the number of leaks and the resultant outage time is dependent on the number of leaks expected. This document presents an estimate of the future reactor water leak frequency and the average outage time charged to each leak. A water leak is defined as any entry of vater into the reactor moderator during operation. Future vater leaks are expected to result primarily from Van Stone flange failures and miscellaneous causes such as mechanical damage, fuel ruptures, and transverse cracks. It is assumed that continuous emphasis will be placed on corrosion monitoring to determine the tubes that must be replaced to prevent leaks because of internal and external corrosion. The number of water leaks is expected to average between 80 and 100 per year in the future. The normal-range of the number of leaks probably will be between 60 and 120 per year as a result of normal variation in the process tube physical condition. The number of outage hours charged per leak is expected to continue …