The Atomic Energy Commission has assigned the Hanford Atomic Products Operation the responsibility of providing the technology and facilities for the interim reprocessing of slightly enriched uranium fuels discharged from power and propulsion reactors. The fuel from the various reactors vary greatly in physical dimensions and in the composition of the core and cladding. For the purposes of chemical reprocessing, the fuels may be divided into the three main categories of aluminum, Zircaloy or stainless steel clad elements. The fuels may be sub-divided by the nature of the core material; e.g. metallic uranium, uranium dioxide, uranium-molybdenum alloys uranium-aluminum alloys, etc.

Design Test Request PR-20 Calandria Characteristics, outlined the need for experimental data concerning the performance of the calandria under transient conditions. Test data was required to confirm that the moderator dump system would drop the level the required 24 inches in less than one second. The original calandria dump chamber design was modified until the criteria was met. This information is recorded in HW-58333, Interim Report, Design Test PR-20, Calandria Characteristics, which lists the drop for the first 24 inches only.

The Fermi age of thermal neutrons in a lattice containing both water and graphite in some sort of weighted average of the age in water and in graphite independently, with correction for volumes of non-moderating elements or voids. The correct weighting function has been in question during design calculations for the NPR. This paper presents a very simple and direct approach to the problem, resulting in a weighting equation which seems to be theoretically sound. Unfortunately, simple theories cannot be guaranteed to give good results in complicated systems; nevertheless, it appears that the weighting dunction derived here is to be preferred over methods involving empirical relationships which are of doubtful validity in the NPR geometry.

For purposes of an earlier consideration of the effects of in-reactor test loops on PRTR operation and program, the design of the high pressure, H20 cooled fuel test loop was assumed to be similar to that for the high pressure, gas cooled loop. Detailed design criteria for the H20 loop have recently been completed so that this assumption is no longer necessary. Therefore, to insure that the effects of all proposed in-reactor loops are fully evaluated with respect to Plutonium Recycle Program objectives, the present study has been carried out. Much of the qualitive discussion in the original analysis is still considered valid. Thus, rather than repeating the entire analysis as it pertains to the H20 loop, only those departures resulting from differences between the assumed design and the scope design are presented.

The Non-Production Fuel Reprocessing Program involves the chemical processing of valuable reactor fuels received from privately owned power reactors. It is necessary therefore, to accurately measure the fuel material received in order to insure proper payment to reactor operator and to provide the Atomic Energy Commission with appropriate accountability data. The volume measurement study described herein was conducted in order to determine the limits of accuracy that could be obtained in measuring relatively large volumes of solution under plant processing conditions.

This report is being submitted to provide information and data on the planning and installation of a new mechanized drawing system which utilizes a new drawing index system and a 35mm microfilm image in a standard IBM card. The communication of engineering information to date has been primarily the full size contact print produced by exposing and developing light sensitive material. This method produces legible copies; but, it has the following disadvantages : (a) the original drawings are fragile, and become dirty and worn through excessive use and handling; (b) the sensitized paper has a limited shelf life; (c) the reproduction cycle is slow because it requires hand feeding and adjustment of the reproduction machine to the condition of the original; (d) the prints are large and unwieldy to handle as working documents; and (e) the filling of originals and copies is slow and expensive. In addition, providing reference prints which are accessible to engineering personnel in the widely separated areas from a central file and reproduction facility is slow, expensive and time consuming.

Presently rotary drum filters are being used for plutonium oxalate slurry filtration in the 234-5 Building. Changing of the filter cloth used on a rotary drum is a time consuming operation which involves ever increasing radiation exposure to maintenance personnel. Consequently, studies were conducted in the 321 Building on a disk type filter which could be adapted for simple and quick ("one-nut") replacement of the filter medium.

At the request of Process Control Development, Chemical Development Operation, statistical assistance was provided by the Operations Research and Synthesis operation in the design and analysis of an experiment to determine how accurately volumes of solution in a large process vessel could be measured. A detailed discussion of the vessel calibration tests, the associated liquid level measuring equipment and the conclusions thereof are presented in HW-62049. The main purpose of this report is to present the precision and accuracy statements obtained, the method by which they were calculated, and to point out the advantages of this method over other "least square" procedures.

Conceptual design studies conducted at HAPO indicate that the improved thermal efficiencies which can be obtained in a high pressure, high temperature, nuclear-electric power plant might reduce unit power costs substantially. The Plutonium Recycle Test Reactor (PRTR) is designed primarily to investigate various facets of plutonium recycle operation. In order to increases the usefulness of the PRTR as an experimental tool, it is desirable to include facilities for testing materials and fuel concepts under a variety of conditions, including the high temperatures and pressures which may be encountered in future water-cooled power reactors.

Report of data for the distributions of elements 90 through 96 between molten salt solutions of their chlorides and molten aluminum (or alloy) at 725 Celsius.

Continued process improvements in the separation plants provide an incentive for economics in waste storage costs by utilizing existing facilities to their maximum capability consistent with the radiological hazards involved. The major improvements have reduced waste volumes resulting in increased fission product concentration and energy potential. Analyses have been made to determine the effect of this change on the integrity of the existing structures.

The PRTR pressure loop is an experimental facility to be installed in the Plutonium Recycle Test Reactor, Building 309, for use in testing the feasibility of reactor cooling systems utilizing high pressure water or steam as the coolant. It will furnish operating experience with such systems and serve to test fuel elements and components designed to operate in such environments.

The region in which Hanford's chemical processing plant areas (separation areas) are located in the central portion of the Hanford Works area on a plateau, shaped like a giant bar, of glaciofluviatile materials. The general locations of these areas is shown in Figure 1. The region studied rises from 200 to 300 feet above the surrounding area and encompasses approximately 40 square miles or somewhat less than 8% of the Hanford reservation (Figure 2). Within the region are located the chemical separation plants and most of the major waste disposal sites.

The calculation of shielding the thickness for plutonium is complicated by the many different energies represented in the gamma radiation emitted during decay of the plutonium isotopes. Dose rate predictions are also frequently confused by gamma from varying content of fission product impurities in the plutonium, as well as other gamma radiation induced through alpha and neutron particle absorption within the source material or its environment. After assumptions are made for these many factors the radiation data for shielding determination is still frequently inadequate because of wide variations in dose rates resulting from self-absorption. The degree if self-shielding is in turn dependent on nature of the plutonium compound, degree of compactness, weight, and over-all geometrical distribution of the source material. By preparing a variety of plutonium samples representing combinations of these varying factors, actual dose rates and gamma spectra, as obtained from them, can then be extrapolated for application to specific situations.

The processing of irradiated fuels from power and propulsion reactors is planned by Hanford Atomic Products Operation as part of the Atomic Energy Commission's interim reprocessing scheme. The several chemical processes proposed for the reprocessing of these fuels may be divided into two categories: first, total dissolution processes which dissolve both fuel element cladding and core material and, second, selective dissolution processes which dissolve either cladding or core material. The Niflex process uses a nitric and hydrofluoric acid solution for the total dissolution of stainless steel clad uranium fuel elements.

This paper discusses a remotely operated system developed to transport (raise and lower) a 1/2 gram radium gamma radiation source for calibration of the HAPO film badges and finger rings. The system employs the rotometer principle for positioning the source for operating purposes. An accurate timer is utilized to assure the desired exposure time.

Property changes to common graphites resulting from reactor irradiation depends strongly on the temperature of exposure. Radiation affects at 30 C and 400 to 500 C are quite well known; however, at intermediate temperatures, results are not well known although a large volume of graphite in most reactors operate in this range. Production test, IP-22A, was reactivated, after a dormant period of many months, to study radiation damage between 30 and 400 C. This report concerns mainly the experimental details and evaluation of the equipment necessary for the controlled temperature irradiation.

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.

On a project of the magnitude and complexity of the Plutonium Recycle Test Reactor is was considered prudent engineering practice to obtain an independant design review by a competent off-site group that had not participated in any way in the development of the design. Accordingly, the Atomic Power Equipment Department of the General Electric Company was requested to make such an engineering review, and an Assistance to Hanford contract authorizing this work was approved by the Atomic Energy Commission in November, 1958.

The Atomic Energy Commission has assigned Hanford Atomic Products Operation the responsibility of reprocessing some of the slightly enriched uranium fuel elements from nuclear power reactors. Some stainless steel clad fuel elements contain sodium or a sodium-potassium alloy as a heat transfer medium between metallic core and outer sheath. The nature of the reaction between water and sodium is well known; however, the reaction between sodium and an oxidizing acid is more energetic and not so well known. The unknown factor of interest is the pressure generated by the reaction between the next transfer medium and the acid used for fuel dissolution prior to solvent extraction. The scouting studies discussed in this report were made to determine hydraulic pressures developed when sodium is exposed to cold concentrated nitric acid beneath the liquid surface.

The NPR design calls for carbon steel to be the major constituent in the reactor coolant piping system. The piping and its associated fittings will, in all likelihood, be exposed to atmospheric weather conditions during the period of reactor construction. This type of exposure causes rusting. From experience gained during the startup of KER Loop 1 it is expected that there will be initially high NPR coolant activity levels. The high activity during the startup of KER Loop 1 was partially caused by the activation of rust that was eroded from pipe walls. Prevention of rusting on the carbon steel prior to its introduction into the coolant system would reduce the initial activity levels.