The effect of finite resistance in the vacuum-tank walls on the longitudinal stability of an intense beam of particles in an accelerator is investigated theoretically. We show that even if the particle frequency is an increasing function of particle energy, the wall resistance can render the beam unstable against longitudinal bunching. In the absence of frequency spread in the unperturbed beam, the instability occurs with a growth rate that is proportional to (N/{sigma}){sup 1/2}, where N is the number of particles in the beam and {sigma} is the conductivity of the surface material. By means of the Vlasov equation a criterion for beam stability is obtained. In the limit of highly conducting walls the criterion involves the frequency spread in the unperturbed beam, the number of particles N, the beam energy, geometrical properties of the accelerator, but not the conductivity {sigma}. A numerical example presented indicates that certain observations of beam behavior in the MURA 40-Mev-electron accelerator may be related to the phenomenon we investigated.

Operation of the present accelerator is summarized and plans for a new injection system and additional machine modernization are described. Research results of both experimentalists and theorists are reviewed in very brief summaries; references are provided. (RWR)

This report, for August 1964 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; weapons manufacturing operation; and safety and security.

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.

The monthly report for the Hanford Laboratories Operation, August 1964. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research, and applied mathematics, and programming operations are discussed.

This document lists all tanks which were active during each month by Waste Farm, which includes all tanks to which additions or subtractions were made. Each type of waste was indicated in each case, along with the nature of transfer and any pertinent remarks. Also reported are gallonage and inches of liquid in each tank.

This interim report details the irradiation testing of co-product fuel-target assemblies in KER loops 3 and 4 which was authorized by PT-IP-645-D. The purpose was to demonstrate and characterize the simultaneous generation of plutonium and tritium during the irradiation of simulated N-Reactor co-product assemblies. Thirteen assemblies, each consisting of a 1.25% enriched outer fuel or driver tube and an inner, coaxial, lithium bearing target rod, were loaded in KER-3 and irradiated to approximately to approximately 1300 MWD/T tube average at N-Reactor conditions. One of the important aspects of the test was the fuel swelling performance in the large, tubular, enriched driver elements. This report describes the fuel swelling behavior of the driver elements.

This report discusses the irradiation testing of thick-walled single-tube, 1.6% enriched test elements (KSE-5`s) in KER loops 1 and 2 to evaluate swelling behavior of uranium at high temperatures. Test charge K-2-21 consisted of four KSE-5`s containing uranium with 400 ppm iron and 800 ppm aluminum. The object of the test irradiation was to establish the swelling performance of the fuel containing the iron-aluminum additions using the N-fuel composition as a reference fuel. This latter fuel has shown good high temperature performance in the British high-temperature gas-cooled reactors, but has not previously been evaluated at ``high`` specific powers in a high-temperature pressurized-water reactor environment.

The final product from the Reduction Operation at the present time is an unalloyed button of reasonably high purity and density. Considerable effort is being directed toward improvement of button purity and surface finish, or appearance. A direct ingoting step is proposed in place of button pickling as a practical means of improving metal quality and appearance. A second ingoting step is proposed to further improve process economics through improved radiographic integrity. Ingoting of buttons immediately following reduction appears to be a more direct and practical approach to definition and improvement of metal purity and improvement of product appearance. The proposed process change places button ingoting within the scope of the Reduction Operation. The final product from the Reduction Operation will then be an ingot of uniform composition (no segregation), reliable density and plutonium content (no slag inclusions or voids). The ingot could be subdivided in this operation, ready for Foundry use; or it could be cast as an ingot of assured quality for off-site shipment.

The hot die sizing (HDS) process is being considered as a replacement for the existing AlSi brazing process. Fuel cores are clad by passing a preheated core-component assembly through a die to bond the outside surface. Concurrently, the assembly is passed over a die plug to bond the internal surface. End bonding is accomplished in a following step by applying heat and pressure to the sized fuel element. Remaining fabrication steps are essentially the same as those used for AlSi brazed fuel. Initial irradiation tests indicated that hot die sized fuels experienced in-reactor growth behavior which differed from that of the matched AlSi fuels in the tests. Presently, two tests are being irradiated in the reactor for determining the effect of end-bonding method, and end-bonding pressure in combination with fuel core geometry, upon fuel element diametral growth during irradiation. Due to the length of time required for irradiation and examination, a third test to determine the effect of core heat treatment and stress relieving has been prepared and is ready for reactor charging. This report summarizes the fabrication of oil and water-quenched cores pretreated by both alpha annealing and AlSi brazing for {open_quotes}Production Test IP-708-A, Evaluation of Oil Quenched Hot …

None

Irradiation tests of aluminum-lithium co-producer target elements are planned for charging in the N-reactor. Although the target elements are subjected to extensive non-destructive testing to assure that defective pieces are not charged, it is necessary to have some knowledge of the failure characteristics of the targets under reactor conditions. To study the possible behavior of in-reactor failures, a series of defected test elements were tested in the high temperature, pressurized TF-9 facility located at 100 KE. The test objectives were: to follow the progress of the failures of target elements with several types of intentional defects, and to run the tests until, on several of the elements, the failure and corrosion had progressed essentially to a point of total destruction of the core. This report describes the intentionally defected elements used in the test and the progressive results of subjecting these defected elements to high temperature water.

The potential of shipping a bucket of metal that has not had sufficient decay time is a common problem to all Reactor Processing Operations. The only method at present to assure shipping adequately decayed metal is by procedural control. This study was performed to determine the feasibility of establishing decay time for irradiated metal in buckets by radiation activity measurements.

This document provides a method for calculating the compensating effects of horizontal control rods in highly shadowed cases encountered in total control administration. The two-dimensional diffusion theory computer code ``9 ANGIE`` was used to calculate the reactivity effects of the horizontal control rod geometry with respect to total control requirements at B, D, F, and DR Reactors. For total control requirements in the center section of B, D, F, and DR Reactors the horizontal control rods will compensate approximately 25% of the buckling increase presently due to enriched fuel at the top and bottom of the pile. If the enrichment is already compensated by adjacent supplementary poison, the inserted horizontal control rods will reduce the center ``B`` section buckling by an amount equivalent to charging 2 PCCF tubes in the center zone. The direct compensating effect of the horizontal control rods with respect to the enriched fuel on the near side of the ``square pile`` remains unchanged.

The Redox acid precycle flowsheet which has been in use since November of 1959 vas adopted to provide neptunium recovery. Although neptunium recovery on this flowsheet has been satisfactory, several problems have been encountered. Among these are: marginal uranium decontamination, increased hexone degradation and increased corrosion of precycle piping and equipment. To circumvent these problems an acid-deficient precycle flowsheet was proposed. In August, 1964, following replacement of the 1A Column and the D-14 backcycle concentrator, the Redox plant was started up on this flowsheet. Neptunium, which left the 2A Column in the waste stream was recovered by processing the waste stream through the 1S Column on an acid flowsheet. The neptunium, extracted in the 1S Column was stripped in the 10 Column and routed to the D-14 backcycle concentrator in the 10W stream. Using this flow pattern it was planned to accumulate neptunium in the D-14 and F-8 vessels-and the 1A, 1S and 10 Columns. Inadequate stripping of neptunium in the 10 Column with resultant neptunium losses to the waste stream necessitated abandoning this flow pattern and the neptunium stream from the 1S Column has been routed to the 2E Column with the 1A Column product stream. Neptunium is presently …

Production Test IP-544-A authorized the irradiation testing of coextruded Zr-2 jacketed, single tube, thick walled, 1.6 percent enriched tubular elements (KSE-5`s) in KER loops 1 and 2 to evaluate swelling behavior at high uranium temperatures. This PT was intended to permit the evaluation of uranium swelling at N-reactor environmental conditions but at considerably higher uranium temperatures than those occurring in N-reactor fuel. Evaluation of the fuel performances at the authorized maximum fuel temperatures in the 550--600 C temperature range is essential in the design and analysis of single tube fuel elements for use in N-reactor. The test was subsequently supplemented to authorize the irradiation of uranium fuel containing up to 3000 ppm alloying agents for the evaluation of swelling behavior at the high uranium temperatures. This report describes the swelling behavior of a charge of 8 KSE-5 fuel elements irradiated in KER loop 1 to appproximately 2500 MWD/T as Test K-1-22. KER Test K-1-22 is similar (except for exposure level) to KER Test K-2-21 in which a charge of 8 KSE-5 fuel elements were irradiated to 1500 MWD/T. In both tests the charges consisted of 4 KSE-5`s containing uranium with low iron and silicon additions (N-reactor fuel composition) and 4 …

The decontamination cycle (HA) at Purex has occasionally responded favorably to additions of waste rework material. The effect, exhibited in increased extraction of plutonium and improved decontamination from fission products has become known as the ``Tonic effect.`` Efforts to find the agent or agents in waste rework which are responsible for this effect have not been successful. Workers at ORNL, principally C. A. Blake, found that various materials decreased the distribution ratios of zirconium-niobium ruthenium activity. This work covered oximes, ketones, and oxalic acid in the Purex HA system. Their work ended with the conclusion that oxalic acid was the most effective of all agents tested. This paper describes the work by Separations Chemistry Laboratory designed to test these and other additives in the HA system at Purex.

The flexibility of N-Reactor will be demonstrated by comparing (in tabular form) the basic 4000 Mw Pu-only, weapons grade case with a 4800 Mw co-product, weapons grade mode including production of Po-210 and PU-238. The competitive position of N-Reactor in the AEC complex will be illustrated in a table summarizing 1964 production studies.

The loss of flow is considered as far as the flow to the inlet hydraulic connector, inlet plugging and water shutoff time. A mockup revealed no vibration of the fuel element bundles and at the low temperatures present there should be no problem of corrosion. Efforts to assure safety with plutonium in the fuel elements are noted. (GHH)

In conjunction with the proposed thorium target element replacement program, hydraulic tests were performed in the Hydraulics Laboratory in 185-D building to determine the flow characteristics of 1.480-inch OD solid thorium loadings in B, C, D, and K fringe tubes. Included are supply curve data for those orifice combinations which are tentatively to be used with the thorium loads in the various fringe tubes. In addition, predicted supply-demand characteristics are presented for a loading of 1.600-inch OD thorium solids in a K-Reactor Zircaloy-2 process tube equipped with a 0.270-inch flat-plate orifice in a X central zone double-orifice retainer union (cage), Drawing H-1-21952.

This report discusses a comparative irradiation test of fuel elements of N-fuel uranium composition at Hanford.

The Purex L-Cell Package is used for organic stripping and concentrating aqueous plutonium solutions from the final Pu solvent extraction decontamination cycle. At the present time the package is used only during times when the gamma activity of the final product is too high to permit routing the solution to N Cell for final ion-exchange purification or when N Cell is inoperative. The package can be used only to concentrate and then route solutions to the PR Room for loadout into PR cans for storage prior to rework via TK-E6. In addition, operating data have shown that the present package has a low boil-off capacity which is insufficient for both current and forecasted plant production rates. At the present time, flowsheet modifications and processing rate reductions are necessary when the package is in operation. Entry into L Cell and contact maintenance are required to maintain the present package which was installed in 1956. Repairs to the package were made in 1962, and information obtained at that time indicates that extensive repairs to the present package will be required in the near future. The purpose of this report is to present the engineering study for the required modifications to the package …

A proposal has been made for ingoting of buttons, without pickling, immediately after breakout from the reduction vessel. This proposed process change could enable the substitution of an ingot of definable, high purity, and good appearance in place of a button as the final product of the Reduction Operation. The change should also achieve economic advantages and decrease radiation exposure. The packaging and storage of the resultant high-grade ingot will require a change of storage container from that presently used to eliminate plastic materials. High-grade plutonium metal corrodes rapidly when packaged in plastic. The ingots, to be prepared in the process test described below, provide an opportunity for testing of a revised, organic-free storage container. The purposes of this process test do not depend on the nature of the buttons processed, hence the tests could be run with unalloyed buttons or with alloyed buttons (RAB), as current FAB needs dictate.

This report describes a study which was made to determine possible changes in mixer placement at H reactor because of the change from an enriched core load to a natural load and to evaluate movement of metal charges 8 inches downstream. Also studied was the effect of short charging on tube life and mixer position. Special consideration was given to the fact that H reactor will shut down in approximately seven months.