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 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.

The following document was done under the auspices of the U.S. Atomic Energy Commission with the intention of analyzing air core cryogenic magnet coils and its usage in fusion research and high energy nuclear physics applications.

"In this paper the single particle motion in the magnetic field created by a circular array of circular current loops is investigated." (p. 3)

Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine the following impurities: Calcium, Carbon, Chromium, Iron, Lithium, Nickel, Oxygen, Potassium, and Zirconium. Spectrographic analysis was used to determine many other impurities. Sodium samples obtained from experimental apparatus operated as part of the work being conducted for Atomics International were analyzed by these methods.

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.

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.

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.

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.

Abstract: "Preliminary design and experimental work aimed toward fabrication of large magnet coils with encapsulated sodium conductors is outlined. Aspects of cooling cycles are discussed along with heat transfer, thermal stability, and the mechanical reinforcement of such coils."

The purpose of the test was to determine the neutron flux level in a BF3 counter well during the second performance of DL-S-225. The measured activity of the irradiation brass bolt was due entirely to the radioactive decay of Zn65. The thermal neutron flux in the BF3 counter well during the second 1000 hour run was calculated to be 2.5 x 10(9) neutrons/cm2 sec.

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.

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.

Abstract: "The effects of a number of ternary additions on the constitution, casting, and fabricating characteristics and the physical properties of aluminum-35 w/o uranium were investigated. Initial investigations were concerned with the effects of 3 w/o ternary additions on the microstructure and press-forging characteristics of the alloy. It was found that additions of this magnitude often introduced extrinsic phases in the alloy. At the 3 w/o level, additions of germanium, silicon, tin or zirconium inhibited the formation of UAl4 and thereby increased the content of the aluminum matrix in the alloy. It was also noted that these additions decreased the pressures required for extruding, and the tin addition also improved the homogeneity of cast shapes. Lead and palladium also improved the homogeneity of the cast material; however, neither of these was an effective inhibitor of UAl4 and free lead was detected in the alloy to which lead had been added at the ternary. From these studies it appears that tin and zirconium are as effective as silicon in enhancing the fabricating characteristics of aluminum-35 w/o uranium alloys, and may prove superior when evaluated on the bases of casting qualities and recycling characteristics."

"The factors of atomic size, ion core potential, and charge, which contribute to the increment in electrical resistivity produced by solutes in metallic solutions, were separated and experimentally evaluated in liquid sodium. The results are expressed in a semi-empirical relationship which defines the increment in resistivity of a monovalent solution as the sum of the fractional difference in atomic volume of solute and solvent and the difference in atomic number of Na and the solutes Li, K, Cs, Rb, Ag, and Au."

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.

Production of new surface area by fracturing of rock behind an explosive-initiated high amplitude wave is probably an important mechanism for energy absorption in the region near n explosion in rock. Explosive-initiated high amplitude pulses in cylinders of three different rocks and two concretes were studied to determine the pulse amplitude need to damage the cylinders. An average high amplitude wave velocity of half the elastic wave velocity was observed in a porous limestone. Elastic theory is not expected to apply to a wave fracturing a rock. Rapid decreases in original free surface velocity of the greywacke rock which was most thoroughly tested indicate that this rock sustained tensile stresses about a hundred times greater than its probably tensile strength for times a long as fifty microseconds.

The following document is an updated note on the process of building an improved nuclear density gauge within the time interval of September 2 to October 1, 1959.

This report discusses the concept of an improved mass gauging technique with a scintillation counter as the radiation detector in order to meet industrial requirements.

Abstract: "A simple solution to Saha's equation has been obtained. The method of solution involves iteration with respect to the electron pressure or concentration and can be applied to the simultaneous calculations of any number of ions. Sample results are given for lithium and aluminum."

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 purpose of the test is to determine the reason for the decreases in the liquid levels of the reference chambers serving the reactor plant remote liquid level indicators. The remote gages and the local liquid level gages agreed closely for the Reactor Plant Component Cooling Water Expansion Tank and the Coolant Discharge and Vent System flash tank and blow-off tank. There appeared to be no loss of water from the D/P cell reference chambers for these two systems over the period of the test. There was no definite indication of leakage from the Valve Operating System reference chamber, however, the differences between remote indicator readings and the sight glass readings are attributed to sticking of the ball check valve on the upper sight glass, inaccuracy of the readings and instrument error. It is recommended that he lower shut-off valve on the upper sight glass be cleaned and reopened prior to reading the sight glass to ensure that the ball check valve is not stuck in the closed position. No lead age was found in any of the level indicating systems.

The following report discusses direct recording of nuclear pulse height data on magnetic tape that represents an inexpensive method of data storage where some degradation of the original energy resolution can be tolerated.

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.