A research and development program aimed at devising processes for the economical recovery of the potentially valuable long-lived fission products from Purex waste has been under wat at Hanford for several years. When this work has begun, the concentrated waste was primarily a nitric acid solution (6 to 10 M HNO3) containing the fission products and relatively small concentrations of iron, sulfate, and other corrosion products. Flowsheets based on classical separation schemes and rather similar to processes used by the Isotopes Division at the AEC's Oak Ridge operation served to separate the desired fission products from one another and from the corrosion products (1,2,3).These separation schemes employed careful step-wise pH adjustment to precipitate first the iron and then to separate the desired fission products from one another. The flowsheets were demonstrated on a pilot-plant scale with full-level plant waste. However, since the earlier work was complete, plant operations have been modified....

The measurements reported here are companion measurements to those reported earlier in HW-63585. The only significant difference between the measurements is that 1.8 w/o enrichment UO2 fuel was used for the first set, and 2.6 w/o enrichment UO2 fuel was used for the measurements described in this report. The new results will be presented graphically, and for completeness, the details of the measurement will be included here as well as in HW-63585.

Four capsules containing Al-1.65 w/o Pu and Al-12 w/o Si-1.65 w/o Pu were charged into the MTR. These capsules will be irradiated to a burnup of 80-100% of the plutonium atoms to determine the stability of the material at high exposures. An additional sixteen capsules containing 5 to 20 w/o Pu in Al and Al-Si have been prepared. Eight are awaiting reactor space and should be charged soon. The remaining eight are being recanned to increase the end gap between the core and the can. Also four capsules containing high density UO2-PuO2 pellets canned in Zircaloy are awaiting assignment of reactor space.

A 19-rod Zircaloy-clad half-length PRTR spike element successfully irradiated to high plutonium burnout at full power in the ETR shoved partial bonding of the core and cladding. Unsatisfactory autoclave films on the Zircaloy cladding of the PRTR Al-Pu fuel elements are delaying final assembly of the first 30 clusters. Further conditioning of the autoclaves and the availability of permanent etching facilities are expected to correct the present difficulties.

Powder metallurgy is rapidly gaining importance as a means of fabricating nuclear fuel elements and other reactor components. It provides a convenient method for forming metals, unusual combinations of metals, and metal-ceramic combinations. The unique features of this technique which make it desirable for nuclear engineering purposes are the following:

Thermal oxidation of graphite in flowing CO2 is being studied at 650 to 850 C, in a single-pass gas system at atmospheric pressure, by observing weight loss rates. The method is used to provide comparative data for candidate reactor graphites. The effects on oxidation rates of graphite purity, structure, coke type, graphitization temperatures and other manufacturing variables are determined. In addition, the effects of gas flow rates and graphite surface to volume ratios are observed.

Considerable interest has developed in the use of ion-exchange in the nuclear energy field in the last decade. Aside from the obvious use of providing demineralized coolant water for reactors, the projected uses of ion-exchange include the recovery of fission products from aquaeous waste streams and the separation and purification of fissionable materials from spent reactor fuels. The latter process may be incidental to the over-all operation, as is the case with the Purex anion exchange facility, or it may be the prime separation process, as may be the case in the recovery of Pu or U from spent power reactor (PRTR) fuel elements.

Basic Studies. It has been reported previously that a reduction of PuO2 to a suboxide does not occur when a powder sample is heated for one hour at 1450 C. To investigate this anomaly, the present hooded facilities were converted from full air flow to an argon atmosphere to prevent oxidation of a possible suboxide. Five grams of PuO2 powder were heated in dry hydrogen to 1500 C for times of one and eight hours. Immediately after discharge, they were mounted and transferred to a helium atmosphere diffractometer hood. The resulting x-ray diffraction pattern consisted only of the single FCC PuO2 phase.

The Zircaloy-a clad, swaged UOa, 19-rod cluster fuel element for the PRTR was designed to use Zircaloy-a wire spirally wrapped around the fuel rods as spacing members. Such use of unbonded, Zircaloy-a spacers introduced the possibility of fretting corrosion. This paper reports preliminary irradiation tests conducted to determine whether or not such corrosions occurs in this fuel element design.

The basic process for fabricating a swaged fuel rod is simple, easy to control and inexpensive. A zircaloy tube is filled with uranium dioxide powder, the ends temporarily plugged and the loaded tube is swaged to compact the UO2 powder to the required density. The swaged rod is then cut to length and counterbored and then end cape are welded into each end. After several tests and inspections, nineteen rods which meet the quality standards are assembled into a single fuel element ready for irradiation.

In an earlier study on criticality conditions for homogenous mixtures, 2/cm^2 was used as the neutron age for all mixtures of water and uranium. At the higher H/U ratios (low uranium concentration), the calculated critical parameters were in good agreement were in good agreement with experimental data. At the low H/U ratios (high uranium concentrations) the calculated critical parameters were smaller than the experimental ones (more conservative from a nuclear safety point view). These results indicated that using 27 cm^2 as the neutron age gives increasingly conservative results as the H/U ratio decreases.

The measurement of a alpha-particle energy has been used by many radiochemical laboratories for the identification and analysis of alpha-active radio nuclides. The use of the total-ionization method for alpha-active radio-nuclides. The use of the total-ionization method for alpha energy in ionization chamber in which the alpha particle loses all its energy in ionization of the chamber gas. Collection of the electrons thus formed generates a voltage pulse across the chamber capacity which is proportional to the alpha particle energy. This pulse is then amplified using a suitable linear amplifier and fed to a pulses as to amplitude; the information is then recorded or stored. Since the pulse amplitude is proportional to the alpha energy lost to the chamber gas, the pulse height analysis can be used to estimate the energy of the alpha particles and in the case of several alpha emitters of different energies, the relative abundance of the alpha emitters can be determined. An alpha energy analyzer system using the ion collection method has been fabricated for use in radiochemical laboratories required to perform a large number of alpha energy determinations. This report describes the operation, maintenance, and application of this alpha energy analyzer system.

A means for decontaminating the primary system of recirculating type reactor is necessary to insure operation and maintenance. This recirculation system can be contaminated by fuel element rupture products and induced corrosion product activities.

Decontamination of stainless steel and carbon steel used in high temperature recirculation systems is currently being studied to obtain an effective and economical decontamination process for use in these systems. This report presents the preliminary investigation process which has demonstrated very effective decontamination and is low in cost.

In conjunction with the fuel element development program at Hanford, it is desired to determine the effects of cladding and core temperatures, cladding thickness, and exposure upon the swelling behavior of unalloyed uranium. To obtain this information, it is proposed to irradiate several fuel rods, clad by coextrusion with Zr-2, in NeK filled stainless steel capsules. The central uranium temperatures are to be monitored by axial thermocouples. Irradiation tests in the MTR and ETR using capsules of similar design have been and are now being conducted. GKH 3-31, a fuel rod, clad with 0.030" Zr-2, operated in the MTR at an average center fuel temperature of 425 C to an exposure of 2100 MWD/T. GKH 3-57, 3-58, and 3-59 are presently undergoing irradiation in the ETR.

Crib evaluation studies showed the soil uptake of plutonium from the Recuplex process (CAW) waste to be low. Preliminary studies of the low soil adsorption of plutonium revealed the presence of a plutonium nitrate anion complex which could be removed by adsorption on a strong base anion exchange resin.

Thermal Resistance of the contact between to core and the jackets or unbonded fuel elements may easily be the largest source of error in core temperature predictions. The object of this work is to improve contact conductance predictions by measuring conductance of the joint between reactor fuel and cladding materials at joint pressures, temperatures and thermal flux levels approaching reactor service conditions.

It is anticipated that neptunium will be recovered in the Purex process by solvent extraction or ion exchange methods as a nitric acid solution of greater than 0.1 g. Np/1 and containing varying amounts of fission products, plutonium, uranium, and thorium, including Th234 (UX1). At the present time this solution is thermally concentrated in the Purex L-cell package to several grams of neptunium per liter. In this operation the solution is contaminated rather badly with plutonium and stainless steel corrosion products. The present specifications are for the neptunium final product to contain less than 0.1 weight percent plutonium, to be relatively free of gross metallic contaminates, and to be low enough in fission product game activity and Th234-Pa234 (UX1-UX2) beta activity to be handled without resorting to remote techniques.

A series of diffusion length measurements were carried out on graphite stacks of various constructions in an attempt to resolve the discrepancies between the graphite diffusion lengths measured in the Hanford reactors and that value measured in the Hanford Standard Pile. It was found that the diffusion length of the graphite in the Hanford reactors is in good agreement with the value for the Hanford Standard Pile when corrections are made for the absorption and scattering of neutrons by the aluminum process tubes, and for neutrons streaming in the channels of the reactors.

Ultrasonic equipment has been developed for nondestructive testing of Hanford fuel elements. The ultrasonic method has replaced the Frost Test for bonding layer inspection in the Hanford canning line, and provides more accurate and reliable results at lower cost. The method has also been adopted to the testing of new fuel elements for which no other method is available.

In order to obtain the most economical utilization of underground storage facilities it is desirable to maintain a running inventory of heat generation and available self concentration in a given tank. Further, it is believed that such knowledge will be helpful in studying underground storage technology. The calculation of fission product heat generation and available self concentration factor in separations waste storage tanks is a complex process. The complexity is increased greatly when material of varying irradiation history, cooling time, etc., is stored at varying production rates. This document presents in tabular form the power generated from the fission product activity associated with one ton of irradiated uranium using the various operating conditions and decay periods of interest for waste storage considerations.

A portable, light weight transistorized alpha, beta, and gamma survey meter was designed and fabricated.

Pressure drops are reported for forced circulation flow of steam-water mixtures in a 23.5 foot long, 1.43 inch ID, 0.1 inch thick, horizontal annulus. The inner surface of the annulus was uniformly heated over a range from 97,000 to 233,000 Btu/hr-ft², exit pressures extended from 100 to 500 psig, and exit steam qualities varied from 0 to 60% by weight. Liquid water entered the annulus and boiling lengths up to 15 feet were investigated. Moreover, the Woods and the Martinelli and Nelson methods of calculating two-phase pressure drop were applied to the experimental conditions, and the deviations between the analytical and the test results are presented.

A nuclear safety indicator has been described. This is an instrument which will make it possible to determine the safety of performing contact maintenance work on certain long columns used in the continuous flow processing of plutonium.