This document is the third in a series of progress report that records investigations of graphite bodies. Along with the report, two appendices are given to describe the different graphite bodies: "Synthetic Binders for Carbon and Graphite" and "High Temperature Physical Properties of Molded Graphites".

From abstract: "A study is made of various processes that occur prior to the pinch effect when an electric field is applied to a deuterium gas. The variables of the problem are percentage of ionization, the electron and ion temperatures, the resistivity of the gas, and the current density."

A method for determining the concentration of Li6 in aqueous solution has been tested using the nuclear reactions Li6 (n,α)H3 and O16 (H3,n)F18. Annihilation 7 radiation of induced 1.87 hour F18 radioactivity was counted with a well-type scintillation counter, and the radioactivity per millimole of lithium was found to be independent of lithium concentration below about 0.2moles/liter. The sensitivity limit for detecting lithium is less than 0.1 micromole (0.0075 micromole Li6).

The purpose of this study was to develop equations for calculating fision product heating in the HRT-CP underflow pot from measured temperatures and to attempt to correlate the rat of solids accumulation in the underflow pot with fission heating and reactor power. Using fission heating data calculated from relating solids accumulation and heating have been tested. In one case an error of no greater than 26% was incurred in the calculation of the total weight of solids collected during chemical plant runs 17-4, 17-5, and 17-6. Further development work will be done on this correlation.

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.

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.

Abstract: The primary coolant of the SM-1 (APPR-1) was analyzed for short-lived and fission product activities. Manganese-56 was found to be the predominant non-fission product nuclide contributing to the short lived activity. Fission products were found in the coolant. It is concluded these fission products originate from a defects in the cladding and from surface contamination of the fuel elements.

Abstract: The results of activity buildup studies in the SM-1 (APPR-1) performed from February 1958 to January 1959 are reported. Data are presented on the extent, nature, and mechanism of the buildup of long-lived gamma emitting nuclides in the reactor primary system. Mathematical equations to describe the activity buildup are derived. Radiation levels after reactor shutdown are presented, as well as the predicted radiation levels at the end of core life.

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.

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.

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.

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.

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 purpose of the survey was to determine the radiation level in the Turbine Generator Services Building, around the Fuel Handling Canal, in the limited access areas of the reactor plant containers and on the boiler container roofs during plant operation. The test was performed with the plant at approximately 100 per cent power with three loops (1A, 1B, and 1C) in service. The radiation levels of the areas surveyed with a few exceptions were essentially background (.03 MR/HR). These exceptions were the Auxiliary Chamber and the limited access areas between the containers. The highest radiation level obtained in the survey was >2500 MR/HR at the reactor chamber walkway.