Six procedures, in stepwise form, that can be used for the determination of plutonium in several forms by controlled-potential coulometric titration are presented. (auth)

ABS>Hot cell demonstration of the Zirflex decladding process coupled with a modified Purex solvent extraction process was completed using specimens of Zircaloy-clad UO/sub 2/ irradiated to levels of 6150-14,600 Mwd/TU. Soluble losses of uranium and plutonium to the decladding solutions were about 0.05%. Centrifugation of the decladding solution is probably necessary to remove up to 1% of the UO/sub 2/ present as fines resulting from the fracture of low (93 to 95%) density pellets; high (96%) density pellets produced few fines. Approximately 5 hours were required to dissolve the UO/sub 2/ core material (14,000 Mwd/TU) in 4M HNO/sub 3/ versus 6 to 7 hours for unirradiated pellets to produce a solvent extraction feed of 100 g U/l and 3M HNO/sub 3/. Gamma decontamination factors for uranium in the Purex CU stream and plutonium in the BP stream were increased by factors of 2 to 10 from the normal 1.3 x 10/sup 3/ and 2.1 x 10/sup 3/, respectively, by pretreatment of the solvent extraction feed with dincetyl monoxime or its degradation product, oxalic acid. Preliminary data indicate radiation damage degrades the solvent, 30% TBP diluted with Amsco 125- 82, upon one pass through the mixer-settler banks with feed solutions irradiated …

Aqueous methods for recovering uranium from BeO- and Al/sub 2/O/sub 3/- base gas-cooled-reactor fuel elements are being evaluated. Two methods for processing Hastelloy-X--clad pelletized BeO-base fuels containing 60 to 70% UO/ sub 2/, such as the GCRE and MGCR, seem feasible. One method involves mechanical stripping or chopping of the cladding followed by leaching of the uranium from the fuel pellets with boiling 6-l3M HNO/sub 3/. In the other method the cladding and UO/sub 2/ are dissolved in boiling 2M HNO/sub 3/-4M HCl. In either case, most of the BeO matrix remains as an undissolved residue. Pellets containing 70% UO/sub 3/ dissolved completely in less than 20 hr in boiling 8M HNO/sub 3/ containing either 2M H/sub 2/SO/sub 4/ or 0.5M HF, producing solutions containing 4 g of uranium per liter. Fuels of high BeO content, e.g. BeO--5% UO/sub 2/, dissolved only slowly in boiling aqueous reagents. Highest initial rates were in sulfuric acid solutions, log (Rate, mg min/sup -1/cm/sup -2/) = 0.223 (H/sub 2/SO/ sub 4, M) - 2.8l and in HF--NH/sub 4/F solutions. ln boiling 5-8M NH/sub 4/F the initial dissolution rate increased from 0.07 to 3.5 mg min/sup -1/cm/sup -2/ as the HF concentration increased from 0 …

Safe and economic methods of handling radioactive materials off-site are required for the successful operation of nuclear chemical plants. These occasions arise in the shipment of spent fuel, radioactive, isotopes, and liquid wastes. An unsolved problem exists in the development of techniques and sites for the final disposal of waste products. (auth)

The stress analysis performed on the PM-2A reactor vessel and cover is discussed. The maximum combined stress (51,000 psi) occurred in the studs after reaching steady-state conditions. A fatigue analysis indicated that this stress could be safely applied 2500 times, and since the studs do not approach 2500 cycles from initial stud tightening to steady-state conditions, they should not suffer any fatigue damage. (auth)

By equating net radiation to space to I/sup 2/R heating in the SNAP-2 instrument compartments, and neglecting internal radiation between compartments, the panel surface equilibrium temperatures were computed for hot and cold temperature extremes. These extremes are defined by: (1) cold - pre-start phase in the shade, (2) hot - full power phase facing the sun. The results of the heat balances for hottest and coldest cases are presented graphically. These figures relate the equilibrium surface temperatures to the I/sup 2/R heat load dissipated by the panel surface for hot and cold orbits before and after startup. Included as parameters are effective panel area (dependent on Agena-interface design details) and the ..cap alpha../epsilon ratio for the surface coating. If ..cap alpha../epsilon = .3/.9 and A = 1 ft/sup 2/, the extremes of temperatures suffered are from -191/sup 0/F to +212/sup 0/F. This study shows that the normal R/C surface coating (..cap alpha../epsilon = .3/.9) is acceptable regarding allowable maximum surface temperatures, if the effective panel area is not less than 1 ft/sup 2/. It also indicates that further work is warranted regarding internal radiative distribution of heat in order to limit the lowest temperatures to -65/sup 0/F.

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This document describes the objectives, targets and scope of the development programs to be sponsored and conducted by the Irradiation Processing Department in support of the Hanford production reactors during FY-1963 - 1964. it is anticipated that the basic funding level for the IPD Reactor and Metallurgy R&D programs for the next two years will total $6,930,000 annually including overhead. Approximately 70 per cent of these funds will be directed in support of the N Reactor, and the remainder in support of the older reactors. The funding levels for the various reactor and Metallurgy programs requiring technical support are reflected in HW-72790, {open_quotes}Budget Proposals for O2 Program Research and Development, FY-1963 and FY-1964, Irradiation Processing Department{close_quotes}, O. H. Greager, April 23, 1962. This latter document also contains Supplementary Budget Proposals to the Commission for work associated with the Plant Improvement Program for the older reactors and alternate production concedes for the N Reactor, which require funds in addition to the basic program. For current planning purposes, however, the technical efforts described within this Program Document have been restricted to those within the basic funding levels, and if Supplementary Programs are authorized, adjustments will be made accordingly.

This document details activities of the Irradiation Processing Department during the month of November, 1962. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

This is the monthly report for the Hanford Laboratories Operation, November 1962. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

This is the monthly report for the Hanford Laboratories Operation August 1962. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

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The objective of this production test is to evaluate the effect of Zr-2 jacket uniformity, thickness, and temperature on non-uniform clad straining of coextruded Zr-2 jacketed rod samples.

During the normal operation of reactors ionic, particulate and colloidal substances in the coolant are activated as they pass through the neutron flux. The parent substances for the activated materials come from several sources: transient corrosion products from the coolant system piping and fuel element cladding, coolant additives and coolant impurities. Previous studies have shown that a major portion of these substances do not simply pass through the neutron flux field but undergo complex sorption phenomena that result in effluent radio-nuclide levels considerably in excess of what would be expected from activation on a once-through basis. This report is being written to outline studies that will provide activity inventory data that will aid in defining radionuclide retention reactions in reactor coolant systems.

Normal Pile exposure variations and day-to-day operating reactivity adjustments must be compensated by a combination of in-reactor poison and enrichment. The calculation of current reactivity status and the prediction of future trends is an integral part of the Physicist`s responsibility. This document describes the computer program ``CALHOT,`` which is based upon one-group diffusion theory, and is intended to be used in routine reactivity-enrichment calculations. It offers the advantages over present hand calculations of greater accuracy, more flexibility, and, most important, much less calculational time.

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Based on process development work at ORNL, a facility was designed for demonstration of the pot calcination process with a variety of high-level radioactive wastes. In this facility, operational and control problems associated with an integrated process can be identified and solved, procedures to improve characteristics of the calcine can be studied, and important aspects of calcine temperature distribution and fission product behavior can be observed. Installation of the facility is planned for the Hanford Atomic Products Operation. The facility can be installed in any cell or cells having certain basic features. (auth)

Separate abstracts have been prepared for four sections of this report. The sections deal with reactor engineering and physics, nuclear physics, instrument development, and applied mathematics. (N.W.R.)

The designer's concept of a test program for the 30-Mw prototype intermediate heat exchanger and steam generator designed and fabricated as part of the Sodium Components Development Program is presented. The performance data will serve to verify the thermal design, or allow application of improved techniques to future designs, give an improved basis for stress analysis in design of future units, and demonstrate the capability and limitations of the units in relation to the performance specifications for which they were designed. Welding techniques for type 316 stainless steel are described. The specifications and operating conditions of the units are given along with instrumentation drawings showing test equipment design and arrangement. (N.W.R.)

Research and development on metallurglcal aspects of pressurized-water systems is summarized. A survey was made of the methods of determining fuel burnup. The mechanisms and kinetics of the loss of boron during heating at 1135 deg C in various dynamic environments were determined. A model was developed to quantitatively characterize the UO/sup 2/ dispersion microstructure of roll-clad fuel plates relative to an ideal'' dispersion. In order to avoid the loss of boron from UO//sub 2/- stainless steel dispersion fuel plates during fabrication, studies were carried out on a refractory glass containing 4 wt.% B/sub 2/O/sub 3/. By using lowsilicon elemental powder, the undesirable reaction between Eu/sub 2/O/ sub 3/ and Si was eliminated; and 13 full-size SM-1 absorbers were fabricated. Work was continued on the borongradient neutron absorber concept. A design was studied for preparing a composite control rod having an upper section made of a boron-gradient dispersion and the lower tip made of Eu/sub 2/O/sub 3/ and stainless steel. Two fuel elements were examined after significant exposure in SM-1. The examination of the miniature boron-iron samples in the final phase of the MTR irradlation test was performed. Twelve miniature test specimens containing 20, 30, or 40 wt % Eu/sub …

Measurements of thorium resonance-capture activations relative to those of Au/sup 197/, which were made in the Peach Bottom (HTGR) critical assembly, are analyzed and compared with thorium resonance activations calculated from tabulated values of the resonance parameters. The new vanadium-subtraction method that was applied is shown to measure thorium resonance captures relative to those of Au/sup 197/ to within several per cent in this geometry. The measurements show that the lowest several resonances of thorium capture less strongly than would be predicted by the listed parameters. This reduced capture is such that the thin-limit, epicadmium 1/E-spectrum resonance integral is about 84 b instead of the 96 b predicted by listed parameters; the thorium resonance captures for the critical assembly are about 6% less. The approximations currently used in calculating resonance captures in annular lattices probably underestimate captures by several per cent. The 300 deg K resonance-capture rate of ThO/sub 2/ in the annulus is indicated to be several per cent larger than that of metallic thorium, presumably because of a solid-state effect. 46 references. (auth)

Techniques for forming polycrystalline LiF from the melt and for fabricating test specimens were developed and evaluated using single-crystal LiF as a control. Large -grain polycrystalline specimens tested in fourpoint loading always showed some plastic deformation (0.078 to 0.798%) before fracture, but the plastic flow was sharpiy reduced from that of single crystals. An etch was developed revealing dislocations on all crystallographic faces of LiF. Details of plastic deformation in polycrystalline material were investigated. Deformation was inhomogeneous among the grains of an aggregate because of differences in orientation with respect to the applied stress, also within individual grains because of interactions between adjoining grains. Grain boundaries were barriers to slip, but stresses resulting from slip in one grain were transmitted to neighboring grains and often caused local deformation near the boundary. Because of local stresses, local slip systems operated although the resolved shear stresses on them from the applied load were below the critical yield stress. In one case, slip occurred on an (010) plane. Three-grain junctions were areas of high residual stress. Fractures originated at boundaries at or near 3-grain junctions, not as a result of inherent boundary weakness but rather because of high stresses developed at the boundary. …

This document details activities of the Irradiation Processing Department during the month of August, 1962. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; and NPR Project.